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Sample records for nrx-a4-est reactor

  1. Summary of loops in the Chalk River NRX and NRU reactors

    International Nuclear Information System (INIS)

    Nishimura, D.T.

    1980-12-01

    The design and operating parameters of the high pressure, high temperature light water loops in the NRX and NRU reactors are presented to assist experimenters reviewing these facilities for their experiments. The NRX and NRU reactor design and operating data of interest to the experimenters are also presented. (author)

  2. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    1960-08-01

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  3. The poisoning of NRX pile

    International Nuclear Information System (INIS)

    Watson, W.H.

    1959-09-01

    The experimental methods used to study the poisoning of the NRX reactor are described and the operation of the reactor in relation to these methods is reviewed for the period February to September 1948. (author)

  4. Follow-up of CRNL employees involved in the NRX reactor clean-up

    International Nuclear Information System (INIS)

    Werner, M.M.; Myers, D.K.; Morrison, D.P.

    1982-07-01

    Data available to date on the mortality of continuing and retired employees of the Chalk River Nuclear Laboratories are consistent with the Σhealthy workerΣ effect that has been observed in similar studies at other nuclear facilities. Because of an accident at the NRX research reactor in December 1952, the reactor was largely dismantled and rebuilt in 1953-54. These operations involved appreciable radiation exposures to a number of employees. The follow-up of the 850 on-site AECL staff involved in the clean-up has indicated that there were no unusual patterns in the mortality of this group when compared with those of the general population of Ontario

  5. The genesis of NRX - first stop on the road to CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Arsenault, J.E.

    2015-12-15

    The CANDU (Canada Deuterium Uranium) reactor family is traceable to World War Two and to certain principal players. They gave Canada an early entry into the world of practical nuclear energy with the NRX (National Research Experimental) reactor, developed under the auspices of the ABC (America, Britain, Canada) countries. The principal players on the NRX project were, in alphabetical order: James Chadwick, C.D. Howe, Leslie R. Groves and C-J. Mackenzie. Their careers up to about 1940 are outlined in Section 10. (author)

  6. Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

    International Nuclear Information System (INIS)

    Vickerd, Meggan

    2013-01-01

    Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Site characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic

  7. Measurement of the stored energy in the NRX reactor reflector graphite

    Energy Technology Data Exchange (ETDEWEB)

    Hilton, H. B.; Larson, E. A.G.

    1959-07-15

    With the co-operation of workers at Windscale and Harwell, whose assistance is hereby gratefully acknowledged, the stored energy content of the inner reflector graphite of NRX has been measured. Measurements made at three different elevations and at different positions through the reflector show that there is, at present, no danger to NRX from an accidental release of the energy. The energy stored in the reflector in 1958 is less by a factor five to ten than the stored energy as measured in 1953. It appears that there has been a continual release of stored energy since 1954 when, after the rehabilitation, the maximum power was raised to 40 MW. Additional thermocouples have been installed in the inner reflector, and future stored energy measurements are being scheduled. (author)

  8. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90oF

    International Nuclear Information System (INIS)

    Allison, G.M.

    1956-11-01

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95 o F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe 3 O 4 . At pH 7 and with 3-5 ppm of O 2 in the water the corrosion product was a mixture of Fe 3 O 4 and γ-Fe 2 O 3 . At high pH with oxygen present Fe 3 O 4 predominated with some traces of Fe 2 O 3 . The systems tested may he listed in order of increasing corrosiveness: High pH with or without O 2 in the water 2 present and continual purification 2 present and no purification or pH control 2 present. (author)

  9. Research reactor put Canada in the nuclear big time

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    The history of the NRX reactor is briefly recounted. When NRX started up in 1947, it was the most powerful neutron source in the world. It is now the oldest research reactor still operating. NRX had to be rebuilt after an accident in 1952, and its calandria was changed again in 1970. Loops in NRX were used to test fuel for the Nautilus submarine, and the first zircaloy pressure tube in the world. At the present time, NRX is in a 'hot standby' condition as a backup to the NRU reactor, which is used mainly for isotope production. NRX will be decommissioned after completion and startup of the new MAPLE-X reactor

  10. L-Area STS MTR/NRU/NRX Grapple Assembly Closure Mechanics Review

    International Nuclear Information System (INIS)

    Huizenga, D. J.

    2016-01-01

    A review of the closure mechanics associated with the Shielded Transfer System (STS) MTR/NRU/NRX grapple assembly utilized at the Savannah River Site (SRS) was performed. This review was prompted by an operational event which occurred at the Canadian Nuclear Laboratories (CNL) utilizing a DTS-XL grapple assembly which is essentially identical to the STS MTR/NRU/NRX grapple assembly used at the SRS. The CNL operational event occurred when a NRU/NRX fuel basket containing spent nuclear fuel assemblies was inadvertently released by the DTS-XL grapple assembly during a transfer. The SM review of the STS MTR/NRU/NRX grapple assembly will examine the operational aspects of the STS and the engineered features of the STS which prevent such an event at the SRS. The design requirements for the STS NRU/NRX modifications and the overall layout of the STS are provided in other documents.

  11. Some loop experiments in the NRX reactor to study the corrosion of mild steel by flowing water at 90{sup o}F

    Energy Technology Data Exchange (ETDEWEB)

    Allison, G. M.

    1956-11-15

    This work was undertaken to find the water conditions necessary for minimum corrosion in the mild steel thermal shield recirculating systems in NRX and NRU. This report contains the chemical and corrosion results obtained by operating three mild steel loops in which water at 85-95{sup o}F was recirculated through test sections located in J-rod positions in the NRX reactor. Lowest corrosion rates were found when the water was maintained at pH 10.5 with or without oxygen being present. In both cases the corrosion was general in nature and no pitting occurred. At pH 7 with oxygen present in the water severe pitting took place and the corrosion rate was several times higher than similar conditions without oxygen in the water. Under oxygen-free conditions the corrosion product was Fe{sub 3}O{sub 4}. At pH 7 and with 3-5 ppm of O{sub 2} in the water the corrosion product was a mixture of Fe{sub 3}O{sub 4} and {gamma}-Fe{sub 2}O{sub 3}. At high pH with oxygen present Fe{sub 3}O{sub 4} predominated with some traces of Fe{sub 2}O{sub 3}. The systems tested may he listed in order of increasing corrosiveness: High pH with or without O{sub 2} in the water < water at pH 7 with no O{sub 2} present and continual purification < water with no O{sub 2} present and no purification or pH control < water at pH 7 with 3-5 ppm of O{sub 2} present. (author)

  12. Performance of the NRX shut-off rods

    International Nuclear Information System (INIS)

    Manson, R.E.

    1965-08-01

    A new type of shut-off rod of electromechanical design was developed by the American Machine and Foundry Company for use in the NRX reactor following the accident of 1952. The new rods were installed in May, 1956, as part of the control system conversion program which was completed in 1958. Some problems were encountered with limit switch adjustment but minor modifications in design led to much improved operation. he performance of the rods also improved as more experience was gained in the maintenance and adjustment of the various headgear components. Each headgear is now overhauled once a year on a routine basis. The present design of shut-off rod is considered to be very satisfactory. There has only been one occasion when a shut-off rod has failed to come fully down on a trip. Rods have failed to operate correctly on five other occasions but these occurred during shutdown periods or when the reactor was being shutdown manually. (author)

  13. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  14. CRNL research reactor retrofit Emergency Filtration System

    International Nuclear Information System (INIS)

    Philippi, H.M.

    1990-01-01

    This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

  15. Comparison Between Calculated and Measured Cross Section Changes in Natural Uranium Irradiated in NRX

    Energy Technology Data Exchange (ETDEWEB)

    Ahlstroem, P E

    1961-03-15

    It is desirable to obtain an experimental check of the reliability of the methods currently used to determine reactivity changes in a reactor and, with a view to meeting this requirement to some extent, a preliminary comparison has been made between calculated and measured cross-section changes in rods of natural uranium irradiated in NRX. The measurements were made at Harwell in the GLEEP reactor and a description has been given by, inter alia, Ward and Craig. The theory of the calculations, which is briefly described in this report, has been indicated by Littler. The investigation showed that the methods for calculating burn up used at present provides a good illustration of the long-term variations in isotope contents. A satisfactory agreement is obtained with experimental results when calculating apparent cross-section changes in uranium rods due to irradiation if the fission cross- section for {sup 239}Pu is set to 780 b. This is 34 b higher than the figure quoted in BNL - 325 (1958). However, in order to get a good idea as to whether the calculated long-term variations in reactivity really correspond to reality, it is necessary to make further investigations. For this reason the results quoted in this report should be regarded as preliminary.

  16. Fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Feraday, M.A.

    1993-01-01

    This paper includes some statements and remarks concerning the uranium silicide fuels for which there is significant fabrication in AECL, irradiation and defect performance experience; description of two Canadian high flux research reactors which use high enrichment uranium (HEU) and the fuels currently used in these reactors; limited fabrication work done on Al-U alloys to uranium contents as high as 40 wt%. The latter concerns work aimed at AECL fast neutron program. This experience in general terms is applied to the NRX and NRU designs of fuel

  17. Rationalization and future planning for AECL's research reactor capability

    International Nuclear Information System (INIS)

    Slater, J.B.

    1990-01-01

    AECL's research reactor capability has played a crucial role in the development of Canada's nuclear program. All essential concepts for the CANDU reactors were developed and tested in the NRX and NRU reactors, and in parallel, important contributions to basic physics were made. The technical feasibility of advanced fuel cycles and of the organic-cooled option for CANDU reactors were also demonstrated in the two reactors and the WR-1 reactor. In addition, an important and growing radio-isotope production industry was established and marketed on a world-wide basis. In 1984, however, it was recognized that a review and rationalization of the research reactor capability was required. The commercial success of the CANDU reactor system had reduced the scope and size of the required development program. Limited research and development funding and competition from other research facilities and programs, required that the scope be reduced to a support basis essential to maintain strategic capability. Currently, AECL, is part-way through this rationalization program and completion should be attained during 1992/93 when the MAPLE reactor is operational and decisions on NRX decommissioning will be made. A companion paper describes some of the unique operational and maintenance problems which have resulted from this program and the solutions which have been developed. Future planning must recognize the age of the NRU reactor (currently 32 years) and the need to plan for eventual replacement. Strategy is being developed and supporting studies include a full technical assessment of the NRU reactor and the required age-related upgrading program, evaluation of the performance characteristics and costs of potential future replacement reactors, particularly the advanced MAPLE concept, and opportunities for international co-operation in developing mutually supportive research programs

  18. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  19. Detection of gaseous fission products in water - a method of monitoring fuel sheathing failures

    Energy Technology Data Exchange (ETDEWEB)

    Tunnicliffe, P. R.; Whittier, A. C.

    1959-05-15

    The gaseous activities stripped from samples of effluent coolant from the NRU fuel elements tested in the central thimble of the NRX reactor (NRU loop) and from the NRX main effluent have been investigated. The activities obtained from the NRU loop can be attributed to gaseous fission products only. Design data have been obtained for a 'Gaseous Fission Product Monitor' to be installed for use with the NRU reactor. It is expected that this monitor will have high sensitivity to activity indicative of an incipient fuel element sheath failure. No qualitative determination of the various gaseous activities obtained from the NRX effluent has been made. A strong component of 25 {+-}1 seconds half-life is not consistent with O-19. Limited information concerning sheath failures in NRX was obtained. Of six failures observed in parallel with the installed delayed neutron monitors, three of these gave pre-warnings and in each case the gaseous fission product monitor showed a substantially greater sensitivity. An experiment in which small samples of uranium, inserted into the NRX reactor, could be exposed at will to a stream of water showed the behaviour of the two types of monitors to be similar. However, a number of signals were detected only by the gaseous fission product monitor. These can be attributed to its sensitivity to relatively long lived fission products. (author)

  20. An analysis of water reactor burnup data with the METHUSELAH II code

    International Nuclear Information System (INIS)

    Floyd, M.; Hicks, D.

    1964-10-01

    The METHUSELAH II code has been used to predict long term reactivity and isotopic changes in the YANKEE, Dresden and NRX reactors. In general it is shown that there is a satisfactory measure of agreement and the first core lives of YANKEE and Dresden appear well predicted. However there are discrepancies in the isotopic composition of the plutonium formed which appear to be correlated with the degree of hardness of the reactor spectrum. It is demonstrated that plausible changes in nuclear data could reduce the discrepancies. (author)

  1. Power auxiliaries and research reactors. Section 3 of Symposium on the peaceful uses of atomic energy in Australia, 1958, held in Sydney, in June 1958

    Energy Technology Data Exchange (ETDEWEB)

    None

    1958-10-15

    The problems of disposing of the large amounts of highly-radioactive waste resulting from a large-scale nuclear power program are reviewed. The Canadian research reactor NRX is discussed. The DIDO reactor is briefly described and operating experience for the first year at high flux is summarized. The core of the High Flux Australian Research Reactor (HIFAR) is described, and some reactivity balance data are given (T.R.H.)

  2. The development of lower enrichment fuels for Canadian research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Belanger, L; Grolway, C M [AECL, Atomic Energy of Canada Limited, Chalk River, ON (Canada); Foo, M T [CRNL, Combustion Engineering Superheater Ltd., Moncton, NB (Canada)

    1983-08-01

    As part of the world wide move to proliferation resistant fuels, new fuels which use reduced enrichment uranium are being developed for use in the NRX and NRU reactors. A fuel consisting of particles of a USiAl alloy dispersed in an Al matrix has been selected for development along with Al-37 wt% U alloy and Al-U{sub 3}O{sub 8} cermet as backup fuels. This report outlines the progress made in the development of the Al-USiAl and Al-37 wt% U. Results show that good quality extruded rods containing either fuel can be made with techniques similar to those used to fabricate the current NRX and NRU fuels. However, the new fuels will be more expensive to make. Although the oxidation behaviour of the Al-USiAl is not as good as that of the Al-U alloys, its corrosion behaviour in high temperature water does not seem much worse. The oxidation and aqueous corrosion of A-37 wt% U are not much different from those of the Al-U alloys currently used. (author)

  3. Reactor loops at Chalk River

    International Nuclear Information System (INIS)

    Sochaski, R.O.

    1962-07-01

    This report describes broadly the nine in-reactor loops, and their components, located in and around the NRX and NRU reactors at Chalk River. First an introduction and general description is given of the loops and their function, supplemented with a table outlining some loop specifications and nine simplified flow sheets, one for each individual loop. The report then proceeds to classify each loop into two categories, the 'main loop circuit' and the 'auxiliary circuit', and descriptions are given of each circuit's components in turn. These components, in part, are comprised of the main loop pumps, the test section, loop heaters, loop coolers, delayed-neutron monitors, surge tank, Dowtherm coolers, loop piping. Here again photographs, drawings and tables are included to provide a clearer understanding of the descriptive literature and to include, in tables, some specifications of the more important components in each loop. (author)

  4. Effects of heat and pressure on the swelling of irradiated uranium. Part I

    Energy Technology Data Exchange (ETDEWEB)

    Churcman, A. T.; Barnes, R. S.; Cottrell, A. H.

    1956-09-15

    Small pieces of a natural uranium fuel bar from the NRX reactor which had been irradiated to 0.3 - 0.4% burn up have been heat treated either in vacuo or at high pressure and changes in their density measured.

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

    International Nuclear Information System (INIS)

    Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

    2014-01-01

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

  6. An estimate of the reactivity of assemblies of NRX fuel elements in light water

    International Nuclear Information System (INIS)

    Jarvis, R.G.

    1960-03-01

    This report contains calculations on the criticality of assemblies of NRX fuel elements in light water. The elements are dealt with in three sections, 'X rods' of natural uranium, enriched elements of U 235 /A1 alloy and enriched elements of Pu/Al alloy. Values of k ∞ and B 2 are provided for two fuel concentrations for each of the two enriched types and for a range of irradiations of the X rods. The calculations for the X rods provide maximum and minimum values of k ∞ . The maximum values for some lattices are a few per cent above unity. Unfortunately, the present experimental evidence does not prove that it is impossible to achieve values of k ∞ greater than unity in lattices of natural uranium in light water. Hence for safety predictions maximum values have been used. The resulting restrictions are not very severe. It is possible to make critical assemblies of the enriched elements, Part (5) contains a set of recommended minimum spacings such that elements of all kinds may safely be mixed in a stack together. There are also predictions of the minimum critical numbers of complete elements or elements cut into slugs. (author)

  7. Fuel recycling and 4. generation reactors

    International Nuclear Information System (INIS)

    Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

    2012-01-01

    The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  9. Submersion-Subcritical Safe Space (S4) reactor

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. The MAPLE-X concept dedicated to the production of radio-isotopes

    International Nuclear Information System (INIS)

    Heeds, W.

    1985-06-01

    MAPLE is a versatile new Canadian multi-purpose research reactor concept that meets the nuclear aspirations of developing countries. It is planned to convert the NRX reactor at Chalk River Nuclear Laboratories into MAPLE-X as a demonstration prototype of this concept and thereafter to dedicate its operation to the production of radio-isotopes. A description of MAPLE-X and details of molybdenum-99 production are given

  11. 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.

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

    International Nuclear Information System (INIS)

    Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

    2010-01-01

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

  13. JENDL-4.0 benchmarking for fission reactor applications

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. Future development in heavy water reactors in Canada

    International Nuclear Information System (INIS)

    Donnelly, J.; Hart, R.G.

    1982-01-01

    1982 marks the 35th anniversary of the start-up of Canada's first research and test reactor, NRX. Its first power reactor has been operating successfully for the past 20 years. With 5,000 MWe of domestic capacity installed, Canada's major CANDU (Canada Deuterium, Uranium) nuclear program has a further 9,500 MWe under construction in Canada for completion by 1990 as well as committed offshore projects in Argentina, Korea and Romania. The CANDU operating record, by any measure of performance, has been outstanding. This performance is largely due to the discipline imposed on the development, design, construction and operation by two fundamental choices: natural uranium and heavy water. The impact of these two choices on availability, fuel utilization, safety and economics is discussed. Future plans call for building on those characteristics which have made CANDU so successful. When time for change comes, current assessments indicate that it will be possible to convert to more efficient advanced fuel cycles without major changes to the basic CANDU design. Primary attention is being focussed on thorium fuel cycles to ensure an abundant and continuing supply of low cost energy for the long term. The resource savings available from these fuel cycles in expanding systems are reviewed and compared with those available from LWR's and Fast Breeders. The results clearly illustrate the versatility of the CANDU reactor. It can benefit from enrichment plants or get along without them. It can complement LWR's or compete with them. It can complement Fast Breeder Reactors or compete with them as well. In the very long term CANDU's and Fast Breeders combined offer the potential of burning all the world's uranium and all the world's thorium. (author)

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

    International Nuclear Information System (INIS)

    Tombakoglu, M.; Cecen, Y.

    2001-01-01

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

  16. RA-0 reactor. New neutronic calculations; Reactor RA-0. Nuevos calculos neutronicos

    Energy Technology Data Exchange (ETDEWEB)

    Rumis, D; Leszczynski, F

    1991-12-31

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core`s interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author). [Espanol] En este trabajo se actualizan los calculos neutronicos realizados para el reactor RA-0, instalado en la Facultad de Ciencias Exactas, Fisicas y Naturales de la Universidad Nacional de Cordoba. Se describen los calculos realizados hasta la fecha y los resultados obtenidos. Las tecnicas incorporadas al calculo de un reactor como el RA-0 permiten predecir en detalle el comportamiento del flujo en el interior del nucleo y en el reflector, lo que sera una importante ayuda en el diseno de experimentos. En particular, el empleo del codigo WIMSD4 para calculos del reactor completo constituye una novedad en las posibles aplicaciones de ese codigo para resolver problemas que se presentan en la practica. (Autor).

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

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  19. 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

  20. Depleted Reactor Analysis With MCNP-4B

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. Crystallization and preliminary X-ray crystallographic analysis of EstE1, a new and thermostable esterase cloned from a metagenomic library

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Jung-Sue [Department of Biology, Yonsei University, Seoul 120-749 (Korea, Republic of); Protein Network Research Center, Yonsei University, Seoul 120-749 (Korea, Republic of); Rhee, Jin-Kyu [Department of Biotechnology, Yonsei University, Seoul 120-749 (Korea, Republic of); Kim, Dong-Uk [Department of Biology, Yonsei University, Seoul 120-749 (Korea, Republic of); Oh, Jong-Won [Department of Biotechnology, Yonsei University, Seoul 120-749 (Korea, Republic of); Cho, Hyun-Soo, E-mail: hscho8@yonsei.ac.kr [Department of Biology, Yonsei University, Seoul 120-749 (Korea, Republic of); Protein Network Research Center, Yonsei University, Seoul 120-749 (Korea, Republic of)

    2006-02-01

    Recombinant EstE1 protein with a histidine tag at the C-terminus was overexpressed in Escherichia coli strain BL21(DE3) and then purified by affinity chromatography. The protein was then crystallized at 290 K by the hanging-drop vapour-diffusion method. EstE1, a new thermostable esterase, was isolated by functional screening of a metagenomic DNA library from thermal environment samples. This enzyme showed activity towards short-chain acyl derivatives of length C4–C6 at a temperature of 303–363 K and displayed a high thermostability above 353 K. EstE1 has 64 and 57% amino-acid sequence similarity to est{sub pc}-encoded carboxylesterase from Pyrobaculum calidifontis and AFEST from Archaeoglobus fulgidus, respectively. The recombinant protein with a histidine tag at the C-terminus was overexpressed in Escherichia coli strain BL21(DE3) and then purified by affinity chromatography. The protein was crystallized at 290 K by the hanging-drop vapour-diffusion method. X-ray diffraction data were collected to 2.3 Å resolution from an EstE1 crystal; the crystal belongs to space group P4{sub 1}2{sub 1}2, with unit-cell parameters a = b = 73.71, c = 234.23 Å. Assuming the presence of four molecules in the asymmetric unit, the Matthews coefficient V{sub M} is calculated to be 2.2 Å{sup 3} Da{sup −1} and the solvent content is 44.1%.

  2. Nuclear energy. The innovations of the N4 reactor

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

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

  3. Chernobyl: recovery operations and the entombment of Reactor 4

    International Nuclear Information System (INIS)

    Dalziel, S.P.C.

    1988-01-01

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

  4. CNTNAP2 and NRXN1 are mutated in autosomal-recessive Pitt-Hopkins-like mental retardation and determine the level of a common synaptic protein in Drosophila

    DEFF Research Database (Denmark)

    Zweier, Christiane; de Jong, Eiko K; Zweier, Markus

    2009-01-01

    , phenotypically overlapping with Pitt-Hopkins syndrome. With a frequency of at least 1% in our cohort of 179 patients, recessive defects in CNTNAP2 appear to significantly contribute to severe MR. Whereas the established synaptic role of NRXN1 suggests that synaptic defects contribute to the associated...... protein can reorganize synaptic morphology and induce increased density of active zones, the synaptic domains of neurotransmitter release. Moreover, both Nrx-I and Nrx-IV determine the level of the presynaptic active-zone protein bruchpilot, indicating a possible common molecular mechanism in Nrx...

  5. The EL-4 reactor. Changing of a pressure tube on a test loop

    International Nuclear Information System (INIS)

    Foulquier, H.; Clara, P.

    1964-01-01

    Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  7. Reactor power cutback system test experience at YGN 4

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

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

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

    International Nuclear Information System (INIS)

    Andrzejewski, K.; Kulikowska, T.

    1993-12-01

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

  10. School version of ESTE EU

    International Nuclear Information System (INIS)

    Carny, P.; Suchon, D.; Chyly, M.; Smejkalova, E.; Fabova, V.

    2008-01-01

    ESTE EU is information system and software for radiological impacts assessment to the territory of the country in case of radiation accident inside/outside of the country .The program enables to model dispersion of radioactive clouds in small-scale and meso-scale. The system enables the user to estimate prediction of the source term (release to the atmosphere ) for any point of radiation/nuclear accident in Europe (for any point of the release, but especially for the sites of European power reactors ). The system enables to utilize results of real radiological monitoring in the process of source term estimation. Radiological impacts of release to the atmosphere are modelled and calculated across the Europe and displayed in the geographical information system (GIS). The school version of ESTE EU is intended for students of the universities which are interested in or could work in the field of emergency response, radiological and nuclear accidents, dispersion modelling, radiological impacts calculation and urgent or preventive protective measures implementation. The school version of ESTE EU is planned to be donated to specialized departments of faculties in Slovakia, Czech Republic, etc. System can be fully operated in Slovak, Czech or English language. (authors)

  11. School version of ESTE EU

    International Nuclear Information System (INIS)

    Carny, P.; Suchon, D.; Chyly, M.; Smejkalova, E.; Fabova, V.

    2009-01-01

    ESTE EU is information system and software for radiological impacts assessment to the territory of the country in case of radiation accident inside/outside of the country .The program enables to model dispersion of radioactive clouds in small-scale and meso-scale. The system enables the user to estimate prediction of the source term (release to the atmosphere ) for any point of radiation/nuclear accident in Europe (for any point of the release, but especially for the sites of European power reactors ). The system enables to utilize results of real radiological monitoring in the process of source term estimation. Radiological impacts of release to the atmosphere are modelled and calculated across the Europe and displayed in the geographical information system (GIS). The school version of ESTE EU is intended for students of the universities which are interested in or could work in the field of emergency response, radiological and nuclear accidents, dispersion modelling, radiological impacts calculation and urgent or preventive protective measures implementation. The school version of ESTE EU is planned to be donated to specialized departments of faculties in Slovakia, Czech Republic, etc. System can be fully operated in Slovak, Czech or English language. (authors)

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

    Science.gov (United States)

    1992-06-01

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

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

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  14. EST-PAC a web package for EST annotation and protein sequence prediction

    Directory of Open Access Journals (Sweden)

    Strahm Yvan

    2006-10-01

    Full Text Available Abstract With the decreasing cost of DNA sequencing technology and the vast diversity of biological resources, researchers increasingly face the basic challenge of annotating a larger number of expressed sequences tags (EST from a variety of species. This typically consists of a series of repetitive tasks, which should be automated and easy to use. The results of these annotation tasks need to be stored and organized in a consistent way. All these operations should be self-installing, platform independent, easy to customize and amenable to using distributed bioinformatics resources available on the Internet. In order to address these issues, we present EST-PAC a web oriented multi-platform software package for expressed sequences tag (EST annotation. EST-PAC provides a solution for the administration of EST and protein sequence annotations accessible through a web interface. Three aspects of EST annotation are automated: 1 searching local or remote biological databases for sequence similarities using Blast services, 2 predicting protein coding sequence from EST data and, 3 annotating predicted protein sequences with functional domain predictions. In practice, EST-PAC integrates the BLASTALL suite, EST-Scan2 and HMMER in a relational database system accessible through a simple web interface. EST-PAC also takes advantage of the relational database to allow consistent storage, powerful queries of results and, management of the annotation process. The system allows users to customize annotation strategies and provides an open-source data-management environment for research and education in bioinformatics.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

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

  17. Early years of nuclear energy research in Canada

    International Nuclear Information System (INIS)

    Laurence, G.C.

    1980-01-01

    The first experimental attempts in Canada to obtain energy from uranium fission were carried out by the author in the Ottawa laboratories of the National Research Council from 1940 to 42. This program grew into a joint British-Canadian laboratory in Montreal. Work done at this laboratory, which moved to Chalk River in 1946, led to the construction of ZEEP (the first nuclear reactor to operate outside of the United States) NRX, and ultimately to the development of the CANDU power reactors. People involved in the work and events along the way are covered in detail. (LL)

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

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    1992-07-01

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

  20. Fast breeder reactors--lecture 4

    International Nuclear Information System (INIS)

    Marshall, W.; Davies, L.M.

    1986-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  2. Performance of UASB reactors in two stages under different HRT and OLR treating residual waters of swine farming Desempenho de reatores UASB em dois estágios sob diferentes TDH e COV tratando águas residuárias de suinocultura

    Directory of Open Access Journals (Sweden)

    Estevão Urbinati

    2013-04-01

    Full Text Available In this study it was evaluated the effects of hydraulic retention time (HRT and Organic Loading Rate (OLR on the performance of UASB (Upflow Anaerobic Sludge Blanket reactors in two stages treating residual waters of swine farming. The system consisted of two UASB reactors in pilot scale, installed in series, with volumes of 908 and 188 L, for the first and second stages (R1 and R2, respectively. The HRT applied in the system of anaerobic treatment in two stages (R1 + R2 was of 19.3, 29.0 and 57.9 h. The OLR applied in the R1 ranged from 5.5 to 40.1 kg CODtotal (m³ d-1. The average removal efficiencies of chemical oxygen demand (COD and total suspended solids (TSS ranged, respectively, from 66.3 to 88.2% and 62.5 to 89.3% in the R1, and from 85.5 to 95.5% and 76.4 to 96.1% in the system (R1 + R2. The volumetric production of methane in the system (R1 + R2 ranged from 0.295 to 0.721 m³CH4 (m³ reactor d-1. It was found that the OLR applied were not limiting to obtain high efficiencies of CODtotal and TSS removal and methane production. The inclusion of the UASB reactor in the second stage contributed to increase the efficiencies of CODtotal and TSS removal, especially, when the treatment system was submitted to the lowest HRT and the highest OLR.Neste trabalho, avaliou-se o efeito do tempo de detenção hidráulica (TDH e da carga orgânica volumétrica (COV no desempenho de reatores UASB (Upflow Anaerobic Sludge Blanket, em dois estágios, tratando águas residuárias de suinocultura. O sistema foi constituído por dois reatores UASB em escala-piloto, instalados em série, com volumes de 908 e 188 L, para o primeiro e segundo estágios (R1 e R2, respectivamente. Os TDH, aplicados no sistema de tratamento anaeróbio, em dois estágios (R1 + R2, foram de 19,3; 29,0 e 57,9 h. As COVs aplicadas no R1 variaram de 5,5 a 40,1 kg DQOtotal (m³ d-1. As eficiências médias de remoção de demanda química de oxigênio total (DQOtotal e s

  3. Program ELM: A tool for rapid thermal-hydraulic analysis of solid-core nuclear rocket fuel elements

    International Nuclear Information System (INIS)

    Walton, J.T.

    1992-11-01

    This report reviews the state of the art of thermal-hydraulic analysis codes and presents a new code, Program ELM, for analysis of fuel elements. ELM is a concise computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in a nuclear thermal rocket reactor with axial coolant passages. The program was developed as a tool to swiftly evaluate various heat transfer coefficient and friction factor correlations generated for turbulent pipe flow with heat addition which have been used in previous programs. Thus, a consistent comparison of these correlations was performed, as well as a comparison with data from the NRX reactor experiments from the Nuclear Engine for Rocket Vehicle Applications (NERVA) project. This report describes the ELM Program algorithm, input/output, and validation efforts and provides a listing of the code

  4. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2009-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  5. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2008-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  6. Questions about the reactor accident with Chernobyl-4

    International Nuclear Information System (INIS)

    Heijboer, R.J.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Orso, J; Marenzana, A

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Platonov, P.A.

    2007-01-01

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

  9. Pourquoi le Grundgedanke (T. 4.0312 de Wittgenstein est si “Fondamental”?

    Directory of Open Access Journals (Sweden)

    André Leclerc

    1998-12-01

    Full Text Available La pensée fondamentale de Wittgenstein affirme que les constantes logiques ne représentent pas, ne fonctionnent pas sémantiquement comme des noms. En apparence triviale, cette pensée, nous le montrerons, à des ramifications étonnantes dans la philosophie du premier Wittgenstein, en particulier en ce qui concerne sa conception de la logique. Notre but est d'interpréter l'aphorisme en question (T. 4.0312 en montrant comment il se rapporte aux thèses et théories les plus importantes du Tractatus, et comment il permet de les faire tenir ensemble d'une façon cohérente, justifiant par là son caractère "fondamental". Nous tentons également d'apporter une réponse aux questions suivantes : les propositions complexes sont-elles encore - comme les propositions atomiques - des "images" de la réalité ? et quelle est cette "logique des faits" dont parle Wittgenstein dans l'aphorisme sous examen ?

  10. EST2Prot: Mapping EST sequences to proteins

    Directory of Open Access Journals (Sweden)

    Lin David M

    2006-03-01

    Full Text Available Abstract Background EST libraries are used in various biological studies, from microarray experiments to proteomic and genetic screens. These libraries usually contain many uncharacterized ESTs that are typically ignored since they cannot be mapped to known genes. Consequently, new discoveries are possibly overlooked. Results We describe a system (EST2Prot that uses multiple elements to map EST sequences to their corresponding protein products. EST2Prot uses UniGene clusters, substring analysis, information about protein coding regions in existing DNA sequences and protein database searches to detect protein products related to a query EST sequence. Gene Ontology terms, Swiss-Prot keywords, and protein similarity data are used to map the ESTs to functional descriptors. Conclusion EST2Prot extends and significantly enriches the popular UniGene mapping by utilizing multiple relations between known biological entities. It produces a mapping between ESTs and proteins in real-time through a simple web-interface. The system is part of the Biozon database and is accessible at http://biozon.org/tools/est/.

  11. Desempenho de reatores anaeróbios de fluxo ascendente com manta de lodo em dois estágios tratando águas residuárias de suinocultura Performance of two-stage up flow anaerobic sludge blanket reactors treating swine wastewater

    Directory of Open Access Journals (Sweden)

    Adriana M. de Santana

    2005-12-01

    Full Text Available Avaliou-se o desempenho de dois reatores anaeróbios de fluxo ascendente com manta de lodo (UASB em escala-piloto com volumes de 908 L e 188 L, instalados em série, alimentados com águas residuárias de suinocultura com concentrações médias de sólidos suspensos totais (SST variando de 2.216 mg L-1 a 7.131 mg L-1 e submetidos a tempos de detenção hidráulica (TDH de 62,3 e 31,1 h, no primeiro reator, e de 12,9 e 6,5 h, no segundo reator. As eficiências médias de remoção de DQOtotal variaram de 74,0% a 89,6% no Reator 1 e de 34,3% a 45,1% no Reator 2, resultando em valores médios de 86,6% a 93,1% para o sistema de tratamento em dois estágios com carga orgânica volumétrica (COV na faixa de 3,40 a 14,44 kg DQOtotal m-3 reator d-1 no Reator 1. As concentrações de metano no biogás foram acima de 75% para o Reator 1 e de 80% para o Reator 2. Os valores médios de pH variaram na faixa de 6,9 a 8,2 para o efluente do Reator 1 e de 7,0 a 8,3 para o efluente do Reator 2. Os ácidos voláteis totais mantiveram-se estáveis com concentrações médias abaixo de 200 mg L-1. Esses resultados indicaram que as condições de carga orgânica, em termos de DQO e SSV, impostas ao sistema de tratamento anaeróbio em dois estágios, não foram limitantes para que houvesse o desenvolvimento de lodo com microbiota adaptada e com alta atividade, propiciando altas eficiências médias de remoção de matéria orgânica (86,6 a 93,1% para DQOtotal e 85,6 a 88,2% para SSV e taxas de produção de metano de 0,156 a 0,289 m³ CH4 kg-1 de DQO removida.The objective of this work was monitoring the performance of two up flow anaerobic sludge blanket reactors (UASB in a pilot-scale testing with volumes of 908 L and 188 L, installed in series, loaded with swine wastewater with total suspended solids (TSS ranging from 2216 to 7131 mg L-1, submitted to an hydraulic detention time (HDT of 62.3 and 31.1 h, in the first reactor, and 12.9 and 6.5 h, in the second

  12. Transient regimes in a heavy water reactor; Regimes transitoires dans un reacteur a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

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

    1953-07-01

    We studied the variations of power and reactivity of a reactor when we raise in a continuous way the starting plates. During the subcritical regime (negative reactivity), the power is determined by reactivity and by the intensity of the sources of photo neutrons, produced during the previous work of the reactor. When, during the rise of the plates, the reactor, pass by the critical regime (zero reactivity), one notes that the reached power is independent of the initial reactivity. During the sur-critical regime (positive reactivity), the elevation of temperature of the uranium bars slows down the growth of reactivity due to the movements of the plates. The power stretches then toward a value that depends only on the regime of cooling of the reactor and the excess of the available reactivity. This survey permits to choose such a rise speed, that reactivity remains constantly lower to a value beyond which the piloting of the reactor becomes difficult. This result is not more valid, if the intensity of the sources is insufficient, what takes place during the first divergences and after a stop of long length. (author) [French] On etudie les variations de puissance et de reactivite d'un reacteur quand on leve d'une facon continue les plaques de demarrage. Pendant le regime subcritique (reactivite negative), la puissance est determinee par la reactivite et par l'intensite des sources de photoneutrons, produites pendant la marche anterieure du reacteur. Quand, au cours de la montee des plaques, le reacteur passe par le regime critique (reactivite nulle), on constate que la puissance atteinte est independante de la reactivite initiale. Pendant le regime surcritique (reactivite positive), l'elevation de temperature des barres d'uranium ralentit l'accroissement de reactivite due aux mouvements des plaques. La puissance tend alors vers une valeur qui ne depend plus que du regime de refroidissement du reacteur et de l'exces de la reactivite disponible. Cette etude permet de

  13. 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.

  14. 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.

  15. 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

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

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

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

  17. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

  18. EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-02-01

    The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

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

    Science.gov (United States)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1962-03-15

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

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

    International Nuclear Information System (INIS)

    Campioni, Guillaume; Mounier, Claude

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    1984-06-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

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

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2000-01-01

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

  7. The behaviour of zirconium alloys in Santowax OM organic coolant at high temperatures

    International Nuclear Information System (INIS)

    Sawatzky, A.

    1964-10-01

    Zirconium alloys have been exposed to Santowax OM at temperatures of 320 to 400 o C for times as long as 5000 hours. Short-term experiments (less than 2 weeks) were done in stainless-steel bombs and small out-of-pile loops. The X-7 organic loop in the NRX reactor was used to study long-term oxidation and hydriding both in-flux and out-of-flux. The results obtained lead to several tentative conclusions: Aluminum cladding serves as an effective hydrogen barrier; Considerable protection against hydriding is given by zirconium oxide, provided impurities in the organic are carefully controlled; Hydriding is greatly enhanced by the presence of chlorine in the coolant; and, Hydriding is somewhat enhanced by neutron irradiation. Of considerable significance is the fact that a Zircaloy-4 in-reactor test section of the X-7 loop was exposed to Santowax OM at 320 to 400 o C for more than 5000 hours without excessive hydriding. (author)

  8. Estudio hidrodinámico de reactores empacados de flujo ascendente(REFA)

    OpenAIRE

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

    2014-01-01

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

  9. Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

    Directory of Open Access Journals (Sweden)

    Gerardo Moreno

    2012-02-01

    Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

  10. Data exchange between the este systems at the SUJB Prague and at the BMLFUW Vienna in case of emergency

    International Nuclear Information System (INIS)

    Carny, P.; Starostova, V.; Hofer, P.

    2005-01-01

    According to arrangement between the State Office for Nuclear Safety (SUJB) and the Federal Ministry of Agriculture and Forestry, Environment and Water Management, Radiation Protection Division (BMLFUW) there is prepared and regularly tested emergency data exchange between the ESTE systems at the SUJB and at the BMLFUW. The ESTE system is support instrument for off-site emergency response and its main goals in case of severe reactor accidents are: .detection of the way of the release from the reactor core to the environment .detection of the state of the reactor core, .prediction of the source term, .estimation of really observed release rate to the atmosphere, .calculations of radioactive clouds dispersion and radiological impacts assessments. The este systems are implemented at the Emergency Response Centre of the Czech Republic (SUJB) in Prague and Austrian versions are implemented at the Crisis Centre of the Austrian Republic (BMLFUW) in Vienna. The main objective of data exchange according to the above mentioned agreement is to extend bilateral information exchange and make a step in the direction of harmonizing emergency management in case of radiological accidents. Automatic data exchange between the este systems at the SUJB Prague and at the BMLFUW Vienna is performed for testing and training regularly once a month. This assists us to have the system ready in any time. (authors)

  11. Blowing loop in the EL-4 reactor: CO{sub 2} flow control analogue study; Boucle de soufflage de la centrale EL-4 - regulation du debit CO{sub 2} - etude analogique

    Energy Technology Data Exchange (ETDEWEB)

    Chazal, G; Merle, J P; Guillemard, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leroy, C; Robin, L; Jacquin, J C; Cornudet, A [Societe INDATOM, France (France)

    1966-07-01

    This report describes one study which contributed to the construction of the Monts d'Arree nuclear power station: EL-4. The reactor is cooled by a CO{sub 2} current provided by 3 turbo-blower groups. The priming vapour for the turbines is taken at the exit of the main CO{sub 2} - H{sub 2}O exchangers. The operation of EL 4 is based on a high degree of centralization of the controls which attributes an important role to the general regulation circuits. This general regulation includes in particular an internal blowing loop which controls the CO{sub 2} flow. The study of the control of this CO{sub 2} flow is made up of 3 parts: - analogue representation of the reactors cooling circuit and of the turbo blower unit. - first test campaign using the analogue computer describing the natural behaviour of the system in the absence of control. theoretical determination of the regulation factors; definition of the regulation using an analogue computer and second test campaign for recording the performances of the blowing loop. The 4. part of the report deals with the analogue study: analogue equations - development. (authors) [French] Ce rapport prend place parmi les etudes de realisation de la Centrale des Monts d'Arree EL-4. Le reacteur est refroidi par une circulation de CO{sub 2} assuree par 3 groupes turbosoufflantes. La vapeur d'entrainement des turbines est prelevee a la sortie des echangeurs principaux CO{sub 2} - H{sub 2}O. L'exploitation de EL-4 repose sur une centralisation poussee des moyens de controle-commande qui attribue un role essentiel aux circuits de regulation generale. Cette regulation generale comporte en particulier une boucle interne de soufflage qui realise un asservissement du debit de CO{sub 2}. L'etude de cette regulation du debit CO{sub 2} comprend 3 parties: - representation analogique du circuit de refroidissement du reacteur et de l'ensemble turbine-soufflante. - premiere campagne d'essais sur calculateur analogique decrivant le comportement

  12. Construction of the core of the 'heavy water-gas' reactor EL 4; Structures du coeur du reacteur 'eau- lourde-gaz EL 4'

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J L; Foulquier, H; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    problem of thermal insulation around a zirconium alloy liner tube. The neutron absorption equivalent is about 1, 1 mm of Al, and the mean loss around 2 p. 100 of the thermal power of the reactor. The methods proposed have proved practicable as a result of important research and developments on automatic remote control for all the operations which make up the sequences of mounting, demounting and repairing of the construction components. In particular the possibilities opened up by the new techniques of welding tubes from the inside have been extended to other problems connected with the assembling of a reactor. (authors) [French] Le coeur de ce reacteur est constitue par une cuve contenant l'eau lourde, cuve traversee d'une serie de tubes de force dans lesquels circule le gaz caloporteur sous pression de 60 at. Les specifications de depart qui ont joue un role important dans la conception de ces structures concernent des aspects de securite de fonctionnement (chargement du combustible par les deux faces du reacteur, remplacement des structures sur les deux faces du reacteur), des necessites neutroniques (absorption des structures minimum, pas du reseau, diametre des tubes de force) et des considerations thermiques (temperature de sortie 500 C). Ces specifications ont entraine une disposition horizontale des tubes de force et des problemes d'encombrement tres delicats qui ont elimine (pour les dimensions d'EL 4) toute possibilite de recourir a des compensateurs de dilatation sur les tubes de force. II s'ensuit un dessin de cuve semi-rigide dans lequel les tubes de force contribuent pour une part importante a la resistance mecanique de l'ensemble en jouant le role de tirant, d'ou des contraintes elevees sur les jonctions et tubes de force (et le choix des alliages de zirconium). Les structures comprennent le tube de force, les jonctions, l'isolement thermique et le tube de guidage. On expose brievement les moyens d'essais mis en oeuvre et les performances de ces diverses

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

    International Nuclear Information System (INIS)

    2014-01-01

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

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

    Science.gov (United States)

    Kadyk, Thomas; Eikerling, Michael

    2015-08-14

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

  15. Simulación con el código MCNP del reactor nuclear RP-10 en su configuración #14, BOC

    OpenAIRE

    Lázaro, Gerardo; Parreño, Fernando

    2001-01-01

    Se presenta los resultados de exceso de reactividad del núcleo del reactor RP-10 en su configuración 14. Este exceso de reactividad ha sido calculado con MCNP4B con un modelo que describe en detalle las características de los elementos combustibles normales y de control, así como de cada elemento que constituye la configuración de trabajo #14. Este modelo fue previamente utilizado en el reactor RP-0 y ha sido aplicado en la configuración de arranque para el cálculo del exceso de reactividad y...

  16. galaxieEST: addressing EST identity through automated phylogenetic analysis.

    Science.gov (United States)

    Nilsson, R Henrik; Rajashekar, Balaji; Larsson, Karl-Henrik; Ursing, Björn M

    2004-07-05

    Research involving expressed sequence tags (ESTs) is intricately coupled to the existence of large, well-annotated sequence repositories. Comparatively complete and satisfactory annotated public sequence libraries are, however, available only for a limited range of organisms, rendering the absence of sequences and gene structure information a tangible problem for those working with taxa lacking an EST or genome sequencing project. Paralogous genes belonging to the same gene family but distinguished by derived characteristics are particularly prone to misidentification and erroneous annotation; high but incomplete levels of sequence similarity are typically difficult to interpret and have formed the basis of many unsubstantiated assumptions of orthology. In these cases, a phylogenetic study of the query sequence together with the most similar sequences in the database may be of great value to the identification process. In order to facilitate this laborious procedure, a project to employ automated phylogenetic analysis in the identification of ESTs was initiated. galaxieEST is an open source Perl-CGI script package designed to complement traditional similarity-based identification of EST sequences through employment of automated phylogenetic analysis. It uses a series of BLAST runs as a sieve to retrieve nucleotide and protein sequences for inclusion in neighbour joining and parsimony analyses; the output includes the BLAST output, the results of the phylogenetic analyses, and the corresponding multiple alignments. galaxieEST is available as an on-line web service for identification of fungal ESTs and for download / local installation for use with any organism group at http://galaxie.cgb.ki.se/galaxieEST.html. By addressing sequence relatedness in addition to similarity, galaxieEST provides an integrative view on EST origin and identity, which may prove particularly useful in cases where similarity searches return one or more pertinent, but not full, matches and

  17. Nuclear reactor (1960); Reacteurs nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

    Maillard, M L [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires; Leo, M B [Electricite de France (EDF), 75 - Paris (France)

    1960-07-01

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [French] Les premiers reacteurs industriels plutonigenes francais G1 - G2 - G3 du Centre de Marcoule comportent une installation de recuperation d'energie. La production d'electricite de G1 ne compense pas l'energie depensee par ailleurs pour le fonctionnement de l'ensemble, par contre, G2 et G3 doivent fournir chacun une puissance de 25 a 30 MW au reseau national d'Electricite de France. Cette puissance est modeste, mais l'experience acquise grace a ces reacteurs est tres grande et c'est grace a elle qu'il nous sera possible de mettre en exploitation les reacteurs energetiques EDF1 - EDF2 - EDF3. Le memoire decrit comment, avant tout demarrage du reacteur, les essais effectues, en particulier ceux concernant l'installation de recuperation d'energie et le caisson, ont permis d'abreger la phase de montee en puissance. (auteur)

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

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

    2001-09-01

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

  19. Extra heavy oil and refinery residues upgrading through Eni Slurry Technology : first EST commercial unit

    Energy Technology Data Exchange (ETDEWEB)

    Rispoli, G.; Sanfilippo, D.; Amoroso, A [Eni S.p.A., Rome (Italy)

    2009-07-01

    The production of heavy crude oils is projected to continue to grow in the upstream oil industry given that large reserves of unconventional extra heavy crude and bitumen exist in several geographic areas including Canada and Venezuela. As reserves of conventional crude oil continue to decline, these unconventional feedstocks are becoming an opportunity to pursue, but they require effective technologies for upgrading and meeting the growing demand for light and middle distillate fuels. This paper described the proprietary technology that offers a solution to upstream and downstream oil producers for bottom-of-the-barrel upgrading. En i Slurry Technology (EST) is constructing an industrial plant in its Sannazzaro refinery in Italy. The plant is designed to convert 23,000 BPSD of vacuum residue into high quality diesel and other valuable refinery streams such as liquefied petroleum gas, naphtha and jet fuel. EST is an H-addition process characterized by the use of a special homogeneous isothermal intrinsically safe reactor, and of a nano-dispersed non-ageing catalyst. EST converts more than 98 per cent of any type of residues to about 110 per cent volume of light products and distillates or extra heavy oils to high quality bottomless SCO. In typical performance, HDS is greater than 85 per cent, HDM greater than 99 per cent and HDCCR greater than 97 per cent. EST also achieves the target of zero fuel oil - zero coke. 12 refs., 4 tabs., 5 figs.

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

    Science.gov (United States)

    Hoogenboom, J. Eduard; Sjenitzer, Bart L.

    2014-06-01

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

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

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    2013-01-01

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

  2. Sensory regulation of neuroligins and neurexin I in the honeybee brain.

    Directory of Open Access Journals (Sweden)

    Sunita Biswas

    2010-02-01

    Full Text Available Neurexins and neuroligins, which have recently been associated with neurological disorders such as autism in humans, are highly conserved adhesive proteins found on synaptic membranes of neurons. These binding partners produce a trans-synaptic bridge that facilitates maturation and specification of synapses. It is believed that there exists an optimal spatio-temporal code of neurexin and neuroligin interactions that guide synapse formation in the postnatal developing brain. Therefore, we investigated whether neuroligins and neurexin are differentially regulated by sensory input using a behavioural model system with an advanced capacity for sensory processing, learning and memory, the honeybee.Whole brain expression levels of neuroligin 1-5 (NLG1-5 and neurexin I (NrxI were estimated by qRT-PCR analysis in three different behavioural paradigms: sensory deprivation, associative scent learning, and lateralised sensory input. Sensory deprived bees had a lower level of NLG1 expression, but a generally increased level of NLG2-5 and NrxI expression compared to hive bees. Bees that had undergone associative scent training had significantly increased levels of NrxI, NLG1 and NLG3 expression compared to untrained control bees. Bees that had lateralised sensory input after antennal amputation showed a specific increase in NLG1 expression compared to control bees, which only happened over time.Our results suggest that (1 there is a lack of synaptic pruning during sensory deprivation; (2 NLG1 expression increases with sensory stimulation; (3 concomitant changes in gene expression suggests NrxI interacts with all neuroligins; (4 there is evidence for synaptic compensation after lateralised injury.

  3. Present status of fusion reactor materials, 4

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  6. 9 CFR 75.4 - Interstate movement of equine infectious anemia reactors and approval of laboratories, diagnostic...

    Science.gov (United States)

    2010-01-01

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

  7. Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

    International Nuclear Information System (INIS)

    Fuster, S.; Tarabella, A.

    1967-01-01

    The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  9. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J; Millot, J P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une description sommaire de la pile et decrit le

  10. Informal presentations by fuel fabricators and others [contributed by W. Ross, U.S. NRC

    International Nuclear Information System (INIS)

    Ross, W.

    1993-01-01

    This paper is representing the the attitude of only one of the major offices of NRC. This is Office of Nuclear Reactor Regulations. Under this office regulates all phases, safety and environmental and safeguard aspects, of about 70 power reactors and about 70 non-power reactors. Another portion of NRC is the Office of Nuclear Material Safety and Safeguards. Under this office is regulated the licensing activities of around 600 licensees of all types. Including all licenses of transportation, import-export, and such as this, as well as special nuclear materials not related to the use of reactors. So when we talk about regulating the use and the protection of special nuclear materials, we are talking about all types of special nuclear materials, not just the type that is used in a reactor. Consequently, when a review is being made, such as now being made at the direction of the Commission to see if relaxation of regulations can be given to reduced enrichment, it is not only under the auspices of the Office of nuclear reactor regulations, but also of this other office. Therefore, it has to be more or less a policy statement of the Commission. As far as current regulations are concerned, the NRC has one threshold: it's either above 20% enrichment or it's below 20% enrichment. We are going to look and see if the SNM license in this area as well as the SNM license in that area can be changed so that some intermediate regulations can be applied to the steps in between 90% and 20%. We also regulate on the basis of the kilogram formula quantity. Most of our most stringent regulations are based on te possession of kilograms of highly enriched uranium. If uranium enrichment is dropped to the 45-50% level, then the composition of the alloys would have to increase from 21 wt.% (NRU) or 28 wt.% (NRX) to somewhere in region of 40 to 50%, possibly more. We know that the Al-40 wt.% U alloy can be extruded to the required diameter. We also know that the alloy is more brittle than the

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

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-15

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

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

    Science.gov (United States)

    Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

    2011-03-15

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

  14. ESAP plus: a web-based server for EST-SSR marker development.

    Science.gov (United States)

    Ponyared, Piyarat; Ponsawat, Jiradej; Tongsima, Sissades; Seresangtakul, Pusadee; Akkasaeng, Chutipong; Tantisuwichwong, Nathpapat

    2016-12-22

    Simple sequence repeats (SSRs) have become widely used as molecular markers in plant genetic studies due to their abundance, high allelic variation at each locus and simplicity to analyze using conventional PCR amplification. To study plants with unknown genome sequence, SSR markers from Expressed Sequence Tags (ESTs), which can be obtained from the plant mRNA (converted to cDNA), must be utilized. With the advent of high-throughput sequencing technology, huge EST sequence data have been generated and are now accessible from many public databases. However, SSR marker identification from a large in-house or public EST collection requires a computational pipeline that makes use of several standard bioinformatic tools to design high quality EST-SSR primers. Some of these computational tools are not users friendly and must be tightly integrated with reference genomic databases. A web-based bioinformatic pipeline, called EST Analysis Pipeline Plus (ESAP Plus), was constructed for assisting researchers to develop SSR markers from a large EST collection. ESAP Plus incorporates several bioinformatic scripts and some useful standard software tools necessary for the four main procedures of EST-SSR marker development, namely 1) pre-processing, 2) clustering and assembly, 3) SSR mining and 4) SSR primer design. The proposed pipeline also provides two alternative steps for reducing EST redundancy and identifying SSR loci. Using public sugarcane ESTs, ESAP Plus automatically executed the aforementioned computational pipeline via a simple web user interface, which was implemented using standard PHP, HTML, CSS and Java scripts. With ESAP Plus, users can upload raw EST data and choose various filtering options and parameters to analyze each of the four main procedures through this web interface. All input EST data and their predicted SSR results will be stored in the ESAP Plus MySQL database. Users will be notified via e-mail when the automatic process is completed and they can

  15. Evaluación del potencial acidogénico para producción de AGV de melaza de la industria azucarera como valorización de este subproducto

    Directory of Open Access Journals (Sweden)

    María Angélica Palomino

    2016-06-01

    Full Text Available Se evaluó el potencial acidogénico de la melaza de la industria azucarera en 4 diferentes OLR (6,02±4,33; 13,96±7,11; 15,81±4,83; 26,94±13,27kgDQO/m3.d en un reactor de fl ujo ascendente, con lodo granular. El sistema no contó con control de pH e inhibición de la fase metanogénica. El reactor operó en continuo durante 148 días. Para evaluar el potencial acidogénico se utilizó el grado de acidifi cación neto (GAn. Los resultados mostraron que durante las tres primeras OLR el %GAn (29,46 ± 13,01; 20,23 ± 13,67; 24,63 ± 19,49 se mantuvo sin diferencias signifi cativas, pero para la mayor OLR el %GAn disminuyó a la tercera parte (10,21 ± 7,14, mientras la concentración de AGV fue la mayor para esta fase (2644,89mgDQO/L, además se avaluó el balance de DQO para cada una de las fases, donde el % de AGV en el efl uente representó el porcentaje orgánico fermentable rápidamente en el efl uente, estos valores indican que con una recirculación interna se podría mejorar el %GAn u obtener otra serie de productos de base biológica para el aprovechamiento de este residuo. En este artículo se utilizó un reactor de fl ujo ascendente como alternativa a los estudiados (CSRT y batch presentando diferentes resultados

  16. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. Guest editors' introduction to the 4th issue of Experimental Software and Toolkits (EST-4)

    NARCIS (Netherlands)

    Brand, van den M.G.J.; Kienle, H.M.; Mens, K.

    2014-01-01

    Experimental software and toolkits play a crucial role in computer science. Elsevier’s Science of Computer Programming special issues on Experimental Software and Toolkits (EST) provide a means for academic tool builders to get more visibility and credit for their work, by publishing a paper along

  19. Digital computer control on Canadian nuclear power plants -experience to date and the future outlook

    International Nuclear Information System (INIS)

    Pearson, A.

    1977-10-01

    This paper discusses the performance of the digital computer control system at Pickering through the years 1973 to 1976. This evaluation is based on a study of the Pickering Generating Station operating records. The paper goes on to explore future computer architectures and the advantages that could accrue from a distributed system approach. Also outlined are the steps being taken to develop these ideas further in the context of two Chalk River projects - REDNET, an advanced data acquisition system being installed to process information from engineering experiments in NRX and NRU reactors, and CRIP, a prototype communications network using cable television technology. (author)

  20. Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

    International Nuclear Information System (INIS)

    DaSilva, R.L.

    1984-09-01

    A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  2. Cryogenic system design for a compact tokamak reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.; Miller, J.R.

    1988-01-01

    The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

  3. A salmonid EST genomic study: genes, duplications, phylogeny and microarrays

    Directory of Open Access Journals (Sweden)

    Brahmbhatt Sonal

    2008-11-01

    Full Text Available Abstract Background Salmonids are of interest because of their relatively recent genome duplication, and their extensive use in wild fisheries and aquaculture. A comprehensive gene list and a comparison of genes in some of the different species provide valuable genomic information for one of the most widely studied groups of fish. Results 298,304 expressed sequence tags (ESTs from Atlantic salmon (69% of the total, 11,664 chinook, 10,813 sockeye, 10,051 brook trout, 10,975 grayling, 8,630 lake whitefish, and 3,624 northern pike ESTs were obtained in this study and have been deposited into the public databases. Contigs were built and putative full-length Atlantic salmon clones have been identified. A database containing ESTs, assemblies, consensus sequences, open reading frames, gene predictions and putative annotation is available. The overall similarity between Atlantic salmon ESTs and those of rainbow trout, chinook, sockeye, brook trout, grayling, lake whitefish, northern pike and rainbow smelt is 93.4, 94.2, 94.6, 94.4, 92.5, 91.7, 89.6, and 86.2% respectively. An analysis of 78 transcript sets show Salmo as a sister group to Oncorhynchus and Salvelinus within Salmoninae, and Thymallinae as a sister group to Salmoninae and Coregoninae within Salmonidae. Extensive gene duplication is consistent with a genome duplication in the common ancestor of salmonids. Using all of the available EST data, a new expanded salmonid cDNA microarray of 32,000 features was created. Cross-species hybridizations to this cDNA microarray indicate that this resource will be useful for studies of all 68 salmonid species. Conclusion An extensive collection and analysis of salmonid RNA putative transcripts indicate that Pacific salmon, Atlantic salmon and charr are 94–96% similar while the more distant whitefish, grayling, pike and smelt are 93, 92, 89 and 86% similar to salmon. The salmonid transcriptome reveals a complex history of gene duplication that is

  4. Neutron noise in nuclear reactors; Le bruit neutronique des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Blaquiere, A. [Institut National des Sciences et Techniques Nucleaires (France); Pachowska, R. [Universite Technique de Varsovie (Poland)

    1961-06-15

    The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

  5. Esplenectomía incidental: ¿Está justificada?

    Directory of Open Access Journals (Sweden)

    Eddy Sierra Enrique

    1997-12-01

    Full Text Available Se revisaron 77 historias clínicas de pacientes que presentaban lesiones incidentales del bazo ocurridas entre 1985 y 1994, durante operaciones intraabdominales y por examen laparoscópico. Las funciones inmunológicas del bazo y su papel contra las infecciones están bien definidas, por lo cual al ser suprimido este órgano a consecuencia de lesiones incidentales, favorecen las infecciones en dichos pacientes de por vida. De las 77 lesiones incidentales se realizaron 47(61 % esplenectomías incidentales y 30(39 % cirugías conservadoras. La úlcera duodenal y la hernia hiatal son las causas principales de esplenectomía incidental. Se presentaron 11 infecciones (23,4 % y 1 fallecido (2,1 % en la esplenectomía incidental contra 1 infección (3,3 % y ningún fallecido en la cirugía conservadora. Se llega a la conclusión de que la esplenectomía no está justificada dadas las desventajas que presenta77 medical histories of patients who presented incidental injuries of the spleen occurred between 1985 and 1994, during intraabdominal operations and by laparoscopic examination, were reviewed. The immunological functions of the spleen and its role against infections are well defined, so the removal of this organ due to incidental injuries favors the appearance of infections in these patients for the rest of their lives. Of the 77 incidental injuries 47 (61 % incidental splenectomies and 30 (39 % conservative surgeries were performed. Duodenal ulcer and hiatal hernia proved to be the main causes of incidental splenectomy. There were 11 infections (23.4 % and 1 death (2.1 % in the incidental splenectomy compared with 1 infection (3.3 % and no death in the conservative surgery. It is concluded that splenectomy is not justified because of its disadvantages

  6. Purification, crystallization and preliminary crystallographic analysis of Est25: a ketoprofen-specific hormone-sensitive lipase

    International Nuclear Information System (INIS)

    Kim, SeungBum; Joo, Sangbum; Yoon, Hyun C.; Ryu, Yeonwoo; Kim, Kyeong Kyu; Kim, T. Doohun

    2007-01-01

    Est25, a ketoprofen-specific hormone-sensitive lipase from a metagenomic library, was crystallized and diffraction data were collected to 1.49 Å resolution. Ketoprofen, a nonsteroidal anti-inflammatory drug, inhibits the synthesis of prostaglandin. A novel hydrolase (Est25) with high ketoprofen specificity has previously been identified using a metagenomic library from environmental samples. Recombinant Est25 protein with a histidine tag at the N-terminus was expressed in Escherichia coli and purified in a homogenous form. Est25 was crystallized from 2.4 M sodium malonate pH 7.0 and X-ray diffraction data were collected to 1.49 Å using synchrotron radiation. The crystals belong to the monoclinic space group C2, with unit-cell parameters a = 197.8, b = 95.2, c = 99.4 Å, β = 97.1°

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

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

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

  8. Notes on a homogeneous reactor project; Idees sur un projet de reacteur homogene

    Energy Technology Data Exchange (ETDEWEB)

    Benveniste, J; Bernot, J; Eidelman, D; Grenon, M; Portes, L; Raspaud, G; Tachon, J; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod, L; Cohen de Lara, G; Delachanal, M; Fontanet, P; Halbronn, G [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France)

    1958-07-01

    An attempt has been made to develop certain ideas concerning homogeneous reactors. The project under consideration is based on the simultaneous use of a suspension of uranium dispersed in heavy or light water and of boiling in the reactor for heat extraction. However, the studies of suspensions and of boiling are relatively independent and can also be developed for reactors of different types using one or the other. Our aim is a minimum investment in fissile material; for this we propose to extract the steam directly from the core and to make use of a cyclone to accelerate this extraction; a cyclone-type circulation creating a field of increasing tangential velocities of the fluid towards the axis causes the droplets of vapour to accelerate towards the axial vortex in which they are collected; the steam output is then evacuated to the external heat utilisation system, for example an exchanger of the condenser-boiler type. The input speed of water into the reactor being one of the important parameters in the running of the pile, a spiral supply input chamber is used, allowing this speed to be regulated in amount and direction. (author)Fren. [French] Nous nous sommes attaches a developper certaines idees relatives aux piles homogenes. Le projet que nous etudions est base sur l'emploi simultane d'une suspension contenant de l'uranium disperse dans l'eau legere ou lourde et de l'ebullition dans le reacteur pour l'extraction de chaleur. Neanmoins, les etudes de suspensions et d'ebullition sont relativement independantes et peuvent egalement etre developpees pour des reacteurs de type different utilisant l'une ou l'autre. Le but que nous cherchons a atteindre est un investissement minimum en matiere fissile; pour cela, nous proposons d'extraire directement la vapeur dans le coeur et de recourir a un dispositif cyclone pour accelerer cette extraction; une circulation type cyclone creant un champ de vitesses tangentielles du fluide croissantes veraxe a pour effet d

  9. AcEST(EST sequences of Adiantum capillus-veneris and their annotation) - AcEST | LSDB Archive [Life Science Database Archive metadata

    Lifescience Database Archive (English)

    Full Text Available List Contact us AcEST AcEST(EST sequences of Adiantum capillus-veneris and their annotation) Data detail Dat...a name AcEST(EST sequences of Adiantum capillus-veneris and their annotation) DOI 10.18908/lsdba.nbdc00839-0...01 Description of data contents EST sequence of Adiantum capillus-veneris and its annotation (clone ID, libr...le search URL http://togodb.biosciencedbc.jp/togodb/view/archive_acest#en Data acquisition method Capillary ...ainst UniProtKB/Swiss-Prot and UniProtKB/TrEMBL databases) Number of data entries Adiantum capillus-veneris

  10. Mortality among long-term Chalk River employees

    International Nuclear Information System (INIS)

    Werner, M.M.; Myers, D.K.

    1986-12-01

    Mortality among Chalk River Nuclear Laboratory (CRNL) employees who died during employment or after retirement has been updated to 1985 December 31. Data in tabular form are presented for overall mortality for male and female employees, for the participants in the clean-up for the NRX and NRU reactor accidents and for a group of CRNL staff with lifetime accumulative doses in excess of 0.2 Sv. Data are also presented on the different types of cancer causing death among male employees. No statistically significant increases in cancer deaths were found in any of the groups analyzed. 25 refs

  11. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    exceptional circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une

  12. Anaerobic Digestion of Sugarcane Vinasse Through a Methanogenic UASB Reactor Followed by a Packed Bed Reactor.

    Science.gov (United States)

    Cabrera-Díaz, A; Pereda-Reyes, I; Oliva-Merencio, D; Lebrero, R; Zaiat, M

    2017-12-01

    The anaerobic treatment of raw vinasse in a combined system consisting in two methanogenic reactors, up-flow anaerobic sludge blanket (UASB) + anaerobic packed bed reactors (APBR), was evaluated. The organic loading rate (OLR) was varied, and the best condition for the combined system was 12.5 kg COD m -3 day -1 with averages of 0.289 m 3 CH 4  kg COD r -1 for the UASB reactor and 4.4 kg COD m -3 day -1 with 0.207 m 3 CH 4  kg COD r -1 for APBR. The OLR played a major role in the emission of H 2 S conducting to relatively stable quality of biogas emitted from the APBR, with H 2 S concentrations <10 mg L -1 . The importance of the sulphate to COD ratio was demonstrated as a result of the low biogas quality recorded at the lowest ratio. It was possible to develop a proper anaerobic digestion of raw vinasse through the combined system with COD removal efficiency of 86.7% and higher CH 4 and a lower H 2 S content in biogas.

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

    International Nuclear Information System (INIS)

    2014-02-01

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

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

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

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

  15. Fusion reactors as a future energy source

    International Nuclear Information System (INIS)

    Seifritz, W.

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    Cognet, G.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-07-01

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

  19. In-pile critical heat flux and post-dryout heat transfer measurements – A historical perspective

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com

    2017-06-15

    In the 1960s’ and 1970s’ Canada was a world leader in performing in-reactor heat transfer experiments on fuel bundles instrumented with miniature sheath thermocouples. Several Critical Heat Flux (CHF) and Post-CHF experiments were performed in Chalk River’s NRU and NRX reactors on water-cooled 3-, 18-, 19-, 21-, and 36-element fuel bundles. Most experiments were obtained at steady-state conditions, where the power was raised gradually from single-phase conditions up to the CHF and beyond. Occasionally, post-dryout temperatures up to 600 °C were maintained for several hours. In some tests, the fuel behaviour during loss-of-flow and blowdown transients was investigated – during these transients sheath temperatures could exceed 2000 °C. Because of the increasingly more stringent licensing requirements for in-pile heat transfer tests on instrumented fuel bundles, no in-pile CHF and post-dryout tests on fuel bundles have been performed anywhere in the world for the past 40 years. This paper provides details of these unique in-pile experiments and describes some of their heat transfer results.

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

    Science.gov (United States)

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

    2013-06-14

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

  1. Safety of research reactors. Topical issues paper no. 4

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-01-30

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

  3. Distributed control and instrumentation systems for future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.

    1976-01-01

    The centralized dual computer system philosophy has evolved as the key concept underlying the highly successful application of direct digital control in CANDU power reactors. After more than a decade, this basis philosophy bears re-examination in the light of advances in system concepts--notably distributed architectures. A number of related experimental programs, all aimed at exploring the prospects of applying distributed systems in Canadian nuclear power plants are discussed. It was realized from the outset that the successful application of distributed systems depends on the availability of a highly reliable, high capacity, low cost communications medium. Accordingly, an experimental facility has been established and experiments have been defined to address such problem areas as interprocess communications, distributed data base design and man/machine interfaces. The design of a first application to be installed at the NRU/NRX research reactors is progressing well

  4. Design options for a bunsen reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  5. Description of the Triton reactor; Pile Triton, rapport descriptif

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-09-01

    The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965. [French] Le reacteur Triton est un reacteur piscine, a uranium enrichi. Il est entre en fonctionnement en 1959, apres une divergence effectuee le 30 juin de cette meme annee. Destine a des etudes de protection contre les rayonnements, son coeur pouvait se deplacer dans le sens longitudinal. La piscine peut etre separee en deux compartiments inegaux par un batardeau. Le coeur triton est place dans le petit compartiment, le coeur Nereide dans le grand compartiment. Un troisieme compartiment sans eau, appele Naiade II, est separe par une paroi en beton dans laquelle est amenagee une fenetre obturee par une plaque d'aluminium (2,50 m x 2,70 m). La fosse Naiade II sert a des experiences de protection utilisant le coeur nereide. Apres une refonte complete, la puissance du reacteur triton qui etait passee progressivement de 1,2 MW a 2 MW, puis 3 MW, a atteint en aout 1965 6, 5 MW. Le

  6. [Project for] a high-flux extracted neutron beam reactor [for physicists]; Un [projet de] reacteur a haut flux et faisceaux sortis [pour physiciens

    Energy Technology Data Exchange (ETDEWEB)

    Ageron, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    French requirements in neutron beams of different energies extracted from a reactor are briefly described. The well-known importance of cold neutrons (above 4 Angstrom) is emphasized. The main characteristics of a reactor suitable for physicists are outlined: They are: 1 - A flux of about 7. 10{sup 14} thermal neutrons in the heavy water of the reflector, 2 - Maximum flexibility obtained by: - physical separation of the core and the reflector, - independence of the different experiments, - possibility of modifying physical experiments up to - and including - the nature of the used reflector, without any appreciable interruption in the operation of the reactor, - reduction of fixed shields to a minimum; ample use of liquid shields (water) and fluid shields (sands). 3 - Technological continuity as far as possible with French research reactors (Siloe, Pegase, Osiris) already existing or under construction. 4 - Safety of operation arising from simplicity of conception. 5 - Minimised construction costs. Lowering of the operating costs is looked for indirectly in the simplification of the solutions and the reduction of operating staff, rather than directly by reducing the consumption of fuel elements and energy. The recommended solution can be described as a closed-core non-pressurized swimming-pool reactor, highly under-moderated by the cooling light water. Surrounding the reactor are a number of 'beam tubes-loops' each consisting of: - a part of the reflector (heavy water in the example described), - a part of neutron extraction beam tube, - the circuits required for their cooling, - the inlet systems of suitable fluids to the beam tube nose (liquid hydrogen in the example described), - the necessary outlets for measurement and control system. The whole 'beam tubes loops' is immersed in the water of the metallic self-supporting swimming-pool. The shielding outside the swimming-pool is composed for the most part by heavy sand in which is the rest of the beam extraction

  7. Low Enrichment Uranium (LEU)-fueled SLOWPOKE-2 nuclear reactor simulation with the Monte-Carlo based MCNP 4A code

    International Nuclear Information System (INIS)

    Pierre, J.R.M.

    1996-01-01

    Following the commissioning of the Low Enrichment Uranium (LEU) Fuelled SLOWPOKE-2 research reactor at the Royal Military College-College Militaire Royal (RMC-CMR), excess reactivity measurements were conducted over a range of temperature and power. The results showed a maximum excess reactivity of 3.37 mk at 33 o C. Several deterministic models using computer codes like WIMS-CRNL, CITATION, TRIVAC and DRAGON have been used to try to reproduce the excess reactivity and temperature trend of both the LEU and HEU SLOWPOKE-2 reactors. The best simulations had been obtained at Ecole Polytechnique de Montreal. They were able to reproduce the temperature trend of their HEU-fuelled reactor using TRIVAC calculations, but this model over-estimated the absolute value of the excess reactivity by 119 mk. Although calculations using DRAGON did not reproduce the temperature trend as well as TRIVAC, these calculations represented a significant improvement on the absolute value at 20 o C reducing the discrepancy to 13 mk. Given the advance in computer technology, a probabilistic approach was tried in this work, using the Monte-Carlo N-Particle Transport Code System MCNP 4A, to model the RMC-CMR SLOWPOKE-2 reactor.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  12. Development of a helical-coil double wall tube steam generator for 4S reactor

    International Nuclear Information System (INIS)

    Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

    2011-01-01

    The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

  13. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

    1987-01-01

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

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

    International Nuclear Information System (INIS)

    1967-01-01

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

  17. The decommissioning of a small nuclear reactor

    International Nuclear Information System (INIS)

    Neset, K.; Christensen, G.C.; Lundby, J.E.; Roenneberg, G.A.

    1990-02-01

    The JEEP II reactor at Kjeller, Norway has been used as a model for a study of the decommissioning of a small research reactor. A radiological survey is given and a plan for volume reducing, packaging, certifying, classifying and shipping of the radioactive waste is described. 23 refs., 4 figs

  18. A history of ZED-2

    International Nuclear Information System (INIS)

    Jones, R.

    2010-01-01

    The ZED-2 Reactor at Chalk River Laboratories was 50 years old this fall. First criticality occurred in September 1960. ZED-2 is perhaps not very well known in the Canadian Nuclear Industry, certainly not as well known as the various CANDU power reactors or the research reactors NRU and NRX. Part of the reason for this I suspect is that when casually judging the importance of reactors the first parameters that spring to mind are power generated (for power reactors) or neutron flux (for research reactors), bigger being 'better in both cases. By these standards ZED-2 does indeed appear puny: the maximum allowed power is 200W and the corresponding flux about 10 8 to 10 9 neutrons cm -2 s -1 , both numbers being about a factor of 500,000 smaller than the corresponding values for NRU. So, what is it all about? How is it that such an apparently insignificant reactor has operated for 50 years?, longer than any other Canadian reactor except NRU and the McMaster Reactor. What is it used for? What contributions has it made to the Canadian industry? Maybe one might also ask for how long is it going to continue? Well, that's what this talk is about, although I think I will leave the final question to wiser heads than mine. ZED-2 is a descendant of famous progenitors: starting with Enrico Farm's first critical pile of graphite and uranium (created at the University of Chicago in 1942) through Canada's ZEEP (first reactor to go critical outside the USA) that went critical in 1945. These early critical facilities were first about proof of principle that a self sustaining nuclear chain reaction could be established and controlled in a reasonable sized facility and second, in the longer term, developing understanding of the underlying reactor physics and the development of theories and methods to accurately predict the important properties of critical assemblies generally. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Campos Gonzalez, Rina Margarita

    2007-07-15

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

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

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

  2. Testing of a reactimeter for a light water reactor in the range + 500 to - 5000 pcm; Essai d'un reactimetre pour reacteur a eau legere dans la gamme + 500, - 5000 pcm

    Energy Technology Data Exchange (ETDEWEB)

    Chauvet, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    This apparatus is designed to measure instantaneously the positive or negative reactivity of a uranium reactor moderated by light water, on condition that the point of departure is the critical state of the reactor, or an already known sub-critical state. Slight modifications only are required to adapt it to another type of reactor. It is an analogue computer which simply inverses the transfer function of the reactor; it is not therefore a model reactor of which the output voltage is connected by a servo-mechanism to the power of the reactor to give the reactivity; the principle of the calculation of the reactivity does not depend on a servomechanism. One of its disadvantages is that it cannot operate outside a power variation range of 2.5 decades. However the measurement of a negative reactivity value between 0 and 3000 pcm is immediate. It measures the reactivity without deducting it from the period; it therefore gives the reactivity very precisely both for divergence and convergence even through in this latter case the period does not in fact exist. The equipment makes it possible to calibrate very rapidly the control rods of a reactor (the rod-drop method), to measure the reactivity of an experiment in the core, and to measure certain temperature effects. It is also possible by introducing a control into the core at a measured rate, to deduce directly its efficiency curve. (author) [French] Cet appareil est destine a mesurer instantanement la reactivite positive ou negative d'un reacteur a uranium modere a l'eau legere, a condition de partir de l'etat critique du reacteur, ou eventuellement d'un etat sous-critique deja connu. De legeres modifications permettent de l'adapter a un autre type de moderateur. C'est un calculateur analogique, qui inverse purement et simplement la fonction de transfert du reacteur; ce n'est donc pas un simulateur de pile dont la tension de sortie est asservie a la puissance du reacteur pour elaborer la reactivite; le principe du

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

    International Nuclear Information System (INIS)

    Motte, F.

    1986-01-01

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

  4. Canadian supercritical water reactor modeling using G4STORK

    International Nuclear Information System (INIS)

    Ford, W.; Buijs, A.

    2015-01-01

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

  5. Canadian supercritical water reactor modeling using G4STORK

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  6. Performance of a multipurpose research electrochemical reactor

    International Nuclear Information System (INIS)

    Henquin, E.R.; Bisang, J.M.

    2011-01-01

    Highlights: → For this reactor configuration the current distribution is uniform. → For this reactor configuration with bipolar connection the leakage current is small. → The mass-transfer conditions are closely uniform along the electrode. → The fluidodynamic behaviour can be represented by the dispersion model. → This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of ±10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  7. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  8. Use of cadmium in solution in the EL 4 reactor moderator irreversible fixing of cadmium on the metallic surfaces; Utilisation du cadmium en solution dans le moderateur du reacteur EL 4 - fixation irreversible du cadmium sur les surfaces metalliques

    Energy Technology Data Exchange (ETDEWEB)

    Croix, O; Paoli, O; Lecomte, J; Dolle, L; Gallic, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the framework of research into the poisoning of the EL-4 reactor by cadmium sulphate, measurements have been made by two different methods of the residual amounts of cadmium liable to be fixed irreversibly on the surfaces in contact with the heavy water. A marked influence of the pH has been noticed. The mechanism of the irreversible fixing is compatible with the hypothesis of an ion-exchange in the surface oxide layer. In a sufficiently wide range of pH the cadmium thus fixed causes very little residual poisoning. The stability of the cadmium sulphate solutions is however rather low in the conditions of poisoning. (authors) [French] Dans le cadre des etudes sur l'empoisonnement du reacteur EL-4 par le sulfate de cadmium, les quantites residuelles de cadmium susceptibles de se fixer irreversiblement sur les parois que mouillerait l'eau lourde, ont ete mesurees experimentalement par deux methodes differentes. On observe une influence nette du pH. Le mecanisme de la fixation irreversible est compatible avec l'hypothese d'un echange d'ions dans la pellicule d'oxyde superficielle. Dans des limites suffisamment larges de pH, la cadmium ainsi fixe n'occasionne pas d'empoisonnement residuel important. La stabilite des solutions de sulfate de cadmium dans les conditions de l'empoisonnement est cependant mediocre. (auteurs)

  9. Deposition of RuO 4 on various surfaces in a nuclear reactor containment

    Science.gov (United States)

    Holm, Joachim; Glänneskog, Henrik; Ekberg, Christian

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

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

    International Nuclear Information System (INIS)

    Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

    1988-04-01

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

  11. 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.)

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

    International Nuclear Information System (INIS)

    1996-09-01

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

  13. Method of fueling for a nuclear reactor

    International Nuclear Information System (INIS)

    Igarashi, Takao.

    1983-01-01

    Purpose: To enable the monitoring of reactor power with sufficient accuracy, upon starting even without existence of neutron source in case of a low average burnup degree in the reactor core. Constitution: Each of fuel assemblies is charged such that neutron source region monitors for the start-up system in a reactor core neutron instrumentation system having nuclear fuel assemblies and a neutron instrumentation system are surrounded with 4 or 16 fuel assemblies of a low burnup degree. Then, the average burnup degree of the fuel assemblies surrounding the neutron source region monitors are increased than the reactor core burnup degree, whereby neutrons released from the peripheral fuels are increased, sufficient number of neutron counts can be obtained even with no neutron sources upon start-up and the reactor power can be monitored at a sufficient accuracy. (Sekiya, K.)

  14. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Science.gov (United States)

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

    2008-12-01

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

  17. A friendly Maple module for one and two group reactor model

    International Nuclear Information System (INIS)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

    2015-01-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  18. A friendly Maple module for one and two group reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O., E-mail: camila.oliv.baptista@gmail.com, E-mail: pavanguilherme@gmail.com, E-mail: kelmo.lins@gmail.com, E-mail: marcelovilelasilva@gmail.com, E-mail: rodrigowerner@hotmail.com, E-mail: neutron201566@yahoo.com, E-mail: vellozo@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  19. CitEST libraries

    Directory of Open Access Journals (Sweden)

    Maria Luísa P. Natividade Targon

    2007-01-01

    Full Text Available In order to obtain a better understanding of what is citrus, 33 cDNA libraries were constructed from different citrus species and genera. Total RNA was extracted from fruits, leaves, flowers, bark, seeds and roots, and subjected or not to different biotic and abiotic stresses (pathogens and drought and at several developmental stages. To identify putative promoter sequences, as well as molecular markers that could be useful for breeding programs, one shotgun library was prepared from sweet orange (Citrus sinensis var. Olimpia. In addition, EST libraries were also constructed for a citrus pathogen, the oomycete Phythophthora parasitica in either virulent or avirulent form. A total of 286,559 cDNA clones from citrus were sequenced from their 5’ end, generating 242,790 valid reads of citrus. A total of 9,504 sequences were produced in the shotgun library and the valid reads were assembled using CAP3. In this procedure, we obtained 1,131 contigs and 4,083 singletons. A total of 19,200 cDNA clones from P. parasitica were sequenced, resulting in 16,400 valid reads. The number of ESTs generated in this project is, to our knowledge, the largest citrus sequence database in the world.

  20. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

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

    International Nuclear Information System (INIS)

    2014-02-01

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

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

    International Nuclear Information System (INIS)

    2007-03-01

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

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

    International Nuclear Information System (INIS)

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-02-15

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

  5. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Zamonsky, G.

    1991-01-01

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author) [es

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

    OpenAIRE

    Hurtado de Mendoza, Diego

    2005-01-01

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

  7. Pairagon+N-SCAN_EST: a model-based gene annotation pipeline

    DEFF Research Database (Denmark)

    Arumugam, Manimozhiyan; Wei, Chaochun; Brown, Randall H

    2006-01-01

    This paper describes Pairagon+N-SCAN_EST, a gene annotation pipeline that uses only native alignments. For each expressed sequence it chooses the best genomic alignment. Systems like ENSEMBL and ExoGean rely on trans alignments, in which expressed sequences are aligned to the genomic loci...... with de novo gene prediction by using N-SCAN_EST. N-SCAN_EST is based on a generalized HMM probability model augmented with a phylogenetic conservation model and EST alignments. It can predict complete transcripts by extending or merging EST alignments, but it can also predict genes in regions without EST...

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

    International Nuclear Information System (INIS)

    Guidez, J.; Saturnin, A.

    2016-01-01

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

  9. Description of the Triton reactor; Pile Triton, rapport descriptif

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-09-01

    The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965. [French] Le reacteur Triton est un reacteur piscine, a uranium enrichi. Il est entre en fonctionnement en 1959, apres une divergence effectuee le 30 juin de cette meme annee. Destine a des etudes de protection contre les rayonnements, son coeur pouvait se deplacer dans le sens longitudinal. La piscine peut etre separee en deux compartiments inegaux par un batardeau. Le coeur triton est place dans le petit compartiment, le coeur Nereide dans le grand compartiment. Un troisieme compartiment sans eau, appele Naiade II, est separe par une paroi en beton dans laquelle est amenagee une fenetre obturee par une plaque d'aluminium (2,50 m x 2,70 m). La fosse Naiade II sert a des experiences de protection utilisant le coeur nereide. Apres une refonte complete, la puissance du reacteur triton qui etait passee progressivement de 1,2 MW a 2 MW, puis 3 MW, a atteint en aout 1965 6, 5 MW

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

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-01-01

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

  11. Crystal structure of hyperthermophilic esterase EstE1 and the relationship between its dimerization and thermostability properties

    Directory of Open Access Journals (Sweden)

    Koh Eunhee

    2007-07-01

    Full Text Available Abstract Background EstE1 is a hyperthermophilic esterase belonging to the hormone-sensitive lipase family and was originally isolated by functional screening of a metagenomic library constructed from a thermal environmental sample. Dimers and oligomers may have been evolutionally selected in thermophiles because intersubunit interactions can confer thermostability on the proteins. The molecular mechanisms of thermostabilization of this extremely thermostable esterase are not well understood due to the lack of structural information. Results Here we report for the first time the 2.1-Å resolution crystal structure of EstE1. The three-dimensional structure of EstE1 exhibits a classic α/β hydrolase fold with a central parallel-stranded beta sheet surrounded by alpha helices on both sides. The residues Ser154, Asp251, and His281 form the catalytic triad motif commonly found in other α/β hydrolases. EstE1 exists as a dimer that is formed by hydrophobic interactions and salt bridges. Circular dichroism spectroscopy and heat inactivation kinetic analysis of EstE1 mutants, which were generated by structure-based site-directed mutagenesis of amino acid residues participating in EstE1 dimerization, revealed that hydrophobic interactions through Val274 and Phe276 on the β8 strand of each monomer play a major role in the dimerization of EstE1. In contrast, the intermolecular salt bridges contribute less significantly to the dimerization and thermostability of EstE1. Conclusion Our results suggest that intermolecular hydrophobic interactions are essential for the hyperthermostability of EstE1. The molecular mechanism that allows EstE1 to endure high temperature will provide guideline for rational design of a thermostable esterase/lipase using the lipolytic enzymes showing structural similarity to EstE1.

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  13. Mo-99 production on a LEU solution reactor

    International Nuclear Information System (INIS)

    Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

    2005-01-01

    A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

  14. Development of EST-derived markers in Dendrobium from EST of related taxa

    OpenAIRE

    Narisa Juejun; Chataporn Chunwongse; Julapark Chunwongse

    2013-01-01

    Public databases are useful for molecular marker development. The major aim of this study was to develop expressedsequence tag (EST)-derived markers in Dendrobium from available ESTs of Phalaenopsis and Dendrobium. A total of 6063sequences were screened for simple sequence repeats (SSRs) and introns. Primers flanking these regions were generated andtested on genomic DNAs of Phalaenopsis and Dendrobium. Twenty-three percent of amplifiable Phalaenopsis EST-derivedmarkers were cross-genera trans...

  15. Magnetite nanoparticles enhance the performance of a combined bioelectrode-UASB reactor for reductive transformation of 2,4-dichloronitrobenzene.

    Science.gov (United States)

    Wang, Caiqin; Ye, Lu; Jin, Jie; Chen, Hui; Xu, Xiangyang; Zhu, Liang

    2017-09-04

    Direct interspecies electron transfer (DIET) among the cometabolism microbes plays a key role in the anaerobic degradation of persistent organic pollutants and stability of anaerobic bioreactor. In this study, the COD removal efficiency increased to 99.0% during the start-up stage in the combined bioelectrode-UASB system (R1) with magnetite nanoparticles addition, which was higher than those in the coupled bioelectrode-UASB (R2; 83.2%) and regular UASB (R3; 71.0%). During the stable stage, the increase of 2,4-dichloronitrobenzene (2,4-DClNB) concentration from 25 mg L -1 to 200 mg L -1 did not affect the COD removal efficiencies in R1 and R2, whereas the performance of R3 was deteriorated obviously. Further intermediates analysis indicated that magnetite nanoparticles enhanced the reductive dechlorination of 2,4-DClNB. High-throughput sequencing results showed that the functional microbes like Syntrophobacter and Syntrophomonas which have been reported to favor the DIET, were predominant on the cathode surface of R1 reactor. It is speculated that the addition of magnetite nanoparticles favors the cooperative metabolism of dechlorinating microbes and electricigens during 2,4-DClNB degradation process in the combined bioelectrode-UASB reactor. This study may provide a new strategy to improve the performance of microbial electrolysis cells and enhance the pollutant removal efficiency.

  16. The accident of Chernobylsk-4 reactor and its consequences

    International Nuclear Information System (INIS)

    1986-01-01

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

  17. ESTIMA, a tool for EST management in a multi-project environment.

    Science.gov (United States)

    Kumar, Charu G; LeDuc, Richard; Gong, George; Roinishivili, Levan; Lewin, Harris A; Liu, Lei

    2004-11-04

    Single-pass, partial sequencing of complementary DNA (cDNA) libraries generates thousands of chromatograms that are processed into high quality expressed sequence tags (ESTs), and then assembled into contigs representative of putative genes. Usually, to be of value, ESTs and contigs must be associated with meaningful annotations, and made available to end-users. A web application, Expressed Sequence Tag Information Management and Annotation (ESTIMA), has been created to meet the EST annotation and data management requirements of multiple high-throughput EST sequencing projects. It is anchored on individual ESTs and organized around different properties of ESTs including chromatograms, base-calling quality scores, structure of assembled transcripts, and multiple sources of comparison to infer functional annotation, Gene Ontology associations, and cDNA library information. ESTIMA consists of a relational database schema and a set of interactive query interfaces. These are integrated with a suite of web-based tools that allow a user to query and retrieve information. Further, query results are interconnected among the various EST properties. ESTIMA has several unique features. Users may run their own EST processing pipeline, search against preferred reference genomes, and use any clustering and assembly algorithm. The ESTIMA database schema is very flexible and accepts output from any EST processing and assembly pipeline. ESTIMA has been used for the management of EST projects of many species, including honeybee (Apis mellifera), cattle (Bos taurus), songbird (Taeniopygia guttata), corn rootworm (Diabrotica vergifera), catfish (Ictalurus punctatus, Ictalurus furcatus), and apple (Malus x domestica). The entire resource may be downloaded and used as is, or readily adapted to fit the unique needs of other cDNA sequencing projects. The scripts used to create the ESTIMA interface are freely available to academic users in an archived format from http

  18. 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.

  19. Concept of transfer functions for a nuclear reactor; Notion de fonction de transfert pour un reacteur nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Dalfes, Abdi [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires. Departement d' Electronique Generale, Service d' Electronique des Reacteurs

    1966-07-01

    The solution to the correlation equations are expressed in terms of the eigenvalues and Eigen-matrices of the transport operator, for a subcritical zero power reactor. This allows to define, for each point of the reactor and for detectors detecting neutrons of given velocities, correlation and transfer functions driven by the same white-noise source. A precise meaning is also given to the importance operator, which is the adjoin of the transport operator. (author) [French] La solution des equations regissant les matrices de correlation est exprimee en fonction des valeurs et matrices propres de l'operateur de transport pour un reacteur sous-critique et de puissance nulle. Ceci permet de definir, en chaque point du reacteur et pour des detecteurs repondant a des neutrons de vitesse definie, des fonctions de correlation et de transfert dont les entrees sont attaquees par une meme source de bruit blanc. Le role joue par l'operateur importance, adjoint de l'operateur de transport, est aussi precise. (auteur)

  20. AcEST: DK948090 [AcEST

    Lifescience Database Archive (English)

    Full Text Available erase A OS=Aspergillus tubingen... 50 1e-05 sp|P0C1S9|DGLB_RAT Sn1-specific diacy...|P47145|YJ77_YEAST Putative lipase YJR107W OS=Saccharomyces ce... 50 8e-06 sp|O42815|FAEA_ASPTU Feruloyl est

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

    International Nuclear Information System (INIS)

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

    1986-07-01

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

  2. Atmospherically dispersed radiocarbon at the Chalk River Laboratories

    International Nuclear Information System (INIS)

    Milton, G.M.; Brown, R.M.; Repta, C.J.W.; Selkirk, C.J.

    1996-01-01

    A small percentage of the total radiocarbon produced by the NRX and NRU experimental reactors at the Chalk River Laboratories has been vented from the main reactor stack and atmospherically dispersed across the site. Surveys conducted in 1982-83 and 1993-94 have shown that atmospheric levels more than 50 m from the stack are never greater than 600 Bq.kg -1 carbon above the natural background level, falling to near-global atmospheric levels at the site boundaries roughly 7 km away. A dispersion factor > 1.2 x 10 6 m 3 .s -1 at ∼ 0.75 km distance from the point of emission is calculated on the basis of recent in-stack monitoring. Analysis of growth rings in on-site trees has provided an opportunity to search for correlations of 14 C output summer power production and/or moderator losses. (author). 16 refs., 14 tabs., 11 figs

  3. Fukushima - calculation of the reactor core inventory and storage pools Dai-ichi 1 to Dai-ichi 4, an estimation of a source term

    International Nuclear Information System (INIS)

    Krpelanova, M.; Carny, P.

    2011-01-01

    Inventory of the reactor core and spent fuel storage pool of the reactors at Dai-ichi 1 to Dai-ichi 4 was determined to need a realistic estimate of the source (released into the atmosphere environment) and modelling of radiological impact of the events in Fukushima NPP. Calculations of inventories were carried out by the methodology that is used in systems to support emergency response and crisis management anymore. Calculations were made based on a model that respects knowledge of real fuels and fuel cycles for individual reactors Dai-ichi. Necessary input data for training the model and calculate inventories are obtained from the IAEA PRIS database.

  4. MDS Nordion - a Canadian success story

    International Nuclear Information System (INIS)

    Boyd, F.

    1998-01-01

    MDS Nordion,the world leading supplier of radioisotopes, had its beginnings as a small division of the crown company Eldorado Mining and Refining Ltd. in the late 1940's. With the end of World War II, Eldorado found itself with a supply of radium left over from the production of uranium during the war and set up a small group in Ottawa to sell it. Most of the stock of radium was disposed of when the development of the NRX reactor made reactor produced radioisotopes possible. The company turned to selling these radioisotopes and soon focused on Cobalt 60 for its use in cancer therapy. In 1952, the group transferred to AECL (Atomic Energy of Canada Ltd.), as part of the Commercial Products Division. In 1988, as part of the federal government's move to privatization, the organization was split into two companies, Theratronics, which manufactures the cancer therapy machines and other equipment for the use of radioisotopes, and Nordion International, which continued the primary business of processing, packaging, selling radioisotopes. In 1991 Nordion International was sold to MDS Inc., a large international group of companies in the life and sciences fields with headquarters in Toronto. The name was changed to MDS Nordion in December 1996

  5. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

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

  6. Aproximaciones a una estética de lo americano

    Directory of Open Access Journals (Sweden)

    María del Milagro Casalla

    2010-01-01

    Full Text Available El legado filosófico que nos ha dejado Rodolfo Kusch tiene cada vez más presencia y valor en este largo itinerario de descubrirnos como americanos. La pasión de este pensador por la América profunda es una fuente de ins- piración para quienes creemos que vivir en América no es simplemente un episodio geográfico sino un horizonte desde y a partir del cual construimos nuestra identidad. En este trabajo se pretende recrear sus consideracio - nes estéticas recogidas en su ensayo “Anotaciones para una estética de lo americano”, escrito en el año 1955. Aunque se trata de uno de los primeros escritos de Rodolfo Kusch, en el mismo se encuentran presentes numerosos conceptos que luego serán tematizados extensamente a lo largo de toda su obra. Considero que el abordaje a sus consideraciones estéticas no ha sido tan frecuentado como sí otros aspectos de su pensamiento (ser-estar, geocultura, símbolo y religión, etc.. Por ello, creo que este trabajo puede constituir una orientación preeliminar para futuras interpretaciones y, a la vez, ser un humilde homenaje a su filosofía, que, a lo largo de los años, sigue estimulando a antiguas y nuevas generaciones de pensadores dedicados a la filosofía latinoamericana.

  7. Anaerobic biogranulation in a hybrid reactor treating phenolic waste

    International Nuclear Information System (INIS)

    Ramakrishnan, Anushyaa; Gupta, S.K.

    2006-01-01

    Granulation was examined in four similar anaerobic hybrid reactors 15.5 L volume (with an effective volume of 13.5 L) during the treatment of synthetic coal wastewater at the mesophilic temperature of 27 ± 5 deg. C. The hybrid reactors are a combination of UASB unit at the lower part and an anaerobic filter at the upper end. Synthetic wastewater with an average chemical oxygen demand (COD) of 2240 mg/L, phenolics concentration of 752 mg/L and a mixture of volatile fatty acids was fed to three hybrid reactors. The fourth reactor, control system, was fed with a wastewater containing sodium acetate and mineral nutrients. Coal waste water contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. A mixture of anaerobic digester sludge and partially granulated sludge (3:1) were used as seed materials for the start up of the reactors. Granules were observed after 45 days of operation of the systems. The granules ranged from 0.4 to 1.2 mm in diameter with good settling characteristics with an SVI of 12 mL/g SS. After granulation, the hybrid reactor performed steadily with phenolics and COD removal efficiencies of 93% and 88%, respectively at volumetric loading rate of 2.24 g COD/L d and hydraulic retention time of 24 h. The removal efficiencies for phenol and m/p-cresols reached 92% and 93% (corresponding to 450.8 and 153 mg/L), while o-cresol was degraded to 88% (corresponding to 51.04 mg/L). Dimethyl phenols could be removed completely at all the organic loadings and did not contribute much to the residual organics. Biodegradation of o-cresol was obtained in the hybrid-UASB reactors

  8. Managing severe reactor accidents. A review and evaluation of our knowledge on reactor accidents and accident management

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

    The report gives a review of the results from the last years research on severe reactor accidents, and an opinion on the possibilities to refine the present strategies for accident management in Swedish and Finnish BWRs. The following aspect of reactor accidents are the major themes of the study: 1. Early pressure relief from hydrogen production; 2. Recriticality in re-flooded, degraded core; 3. Melt-through; 4. Steam explosion after melt-through; 5. Coolability of the melt after after melt-through; 6. Hydrogen fire in the reactor containment; 7. Leaking containment; 8. Hydrogen fire in the reactor building; 9. Long-time developments after a severe accident; 10. Accidents during shutdown for overhaul; 11. Information need for remedial actions. Possibilities for improving the strategies in each of these areas are discussed. The review shows that our knowledge is sufficient in the areas 1, 2, 4, 6, 8. For the other areas, more research is needed

  9. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

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

  10. Nuclear reactor technology progress report, vol. 4

    International Nuclear Information System (INIS)

    1981-01-01

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

  11. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Limaiem, I.

    2006-12-01

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  12. Effect of Co3O4 and CeO2 Infiltration on the Activity of a LSM15/GDC10 Highly Porous Electrochemical Reactor

    DEFF Research Database (Denmark)

    Ippolito, Davide; Kammer Hansen, Kent

    2014-01-01

    The reduction of air pollution has become an international concern over the last ten years because of increases in emissions from mobile and stationary sources. Among these sources, volatile organic compounds (VOC) represent a serious environmental problem, together with NOx, SOx and particulate...... VOC component of Diesel engine exhausts, over a wide range of temperatures. The entire reactor was thought as a highly porous catalytic filter for a possible application in a Diesel exhausts purification system. The porous reactor was used as a backbone for the infiltration of Co3O4 and Co3O4/CeO2...

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

    International Nuclear Information System (INIS)

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

    2009-06-01

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

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

    International Nuclear Information System (INIS)

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

    2012-03-01

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

  15. ESTIMA, a tool for EST management in a multi-project environment

    Directory of Open Access Journals (Sweden)

    Lewin Harris A

    2004-11-01

    Full Text Available Abstract Background Single-pass, partial sequencing of complementary DNA (cDNA libraries generates thousands of chromatograms that are processed into high quality expressed sequence tags (ESTs, and then assembled into contigs representative of putative genes. Usually, to be of value, ESTs and contigs must be associated with meaningful annotations, and made available to end-users. Results A web application, Expressed Sequence Tag Information Management and Annotation (ESTIMA, has been created to meet the EST annotation and data management requirements of multiple high-throughput EST sequencing projects. It is anchored on individual ESTs and organized around different properties of ESTs including chromatograms, base-calling quality scores, structure of assembled transcripts, and multiple sources of comparison to infer functional annotation, Gene Ontology associations, and cDNA library information. ESTIMA consists of a relational database schema and a set of interactive query interfaces. These are integrated with a suite of web-based tools that allow a user to query and retrieve information. Further, query results are interconnected among the various EST properties. ESTIMA has several unique features. Users may run their own EST processing pipeline, search against preferred reference genomes, and use any clustering and assembly algorithm. The ESTIMA database schema is very flexible and accepts output from any EST processing and assembly pipeline. ESTIMA has been used for the management of EST projects of many species, including honeybee (Apis mellifera, cattle (Bos taurus, songbird (Taeniopygia guttata, corn rootworm (Diabrotica vergifera, catfish (Ictalurus punctatus, Ictalurus furcatus, and apple (Malus x domestica. The entire resource may be downloaded and used as is, or readily adapted to fit the unique needs of other cDNA sequencing projects. Conclusions The scripts used to create the ESTIMA interface are freely available to academic users in

  16. A Compact Quasi-axisymmetric Stellarator Reactor

    International Nuclear Information System (INIS)

    Ku, L.P.

    2003-01-01

    We report the progress made in assessing the potential of compact, quasi-axisymmetric stellarators as power-producing reactors. Using an aspect ratio A=4.5 configuration derived from NCSX and optimized with respect to the quasi-axisymmetry and MHD stability in the linear regime as an example, we show that a reactor of 1 GW(e) maybe realizable with a major radius *8 m. This is significantly smaller than the designs of stellarator reactors attempted before. We further show the design of modular coils and discuss the optimization of coil aspect ratios in order to accommodate the blanket for tritium breeding and radiation shielding for coil protection. In addition, we discuss the effects of coil aspect ratio on the peak magnetic field in the coils

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  18. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

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

  19. Evolución del equilibrio estático y dinámico desde los 4 hasta los 74 años

    Directory of Open Access Journals (Sweden)

    Josep Cabedo i Sanromà

    2008-06-01

    Full Text Available La investigación que presentamos proporciona una descripción del equilibrio durante el ciclo vital –desde los 4 hasta los 74 años–, y tiene como objetivo principal el análisis, la valoración y la reflexión, sobre su evolución en dos pruebas: una de equilibrio estático y otra de dinámico. En el apartado de metodología quedan detallados los protocolos de administración y condiciones de observación necesarios para utilizar estos dos tests, por parte de los diferentes profesionales de la actividad física y de otros profesionales que los puedan necesitar. Al mismo tiempo, y con una vocación de ofrecer un servicio social, adjuntamos los baremos por edades de las pruebas de equilibrio estático y dinámico. Los datos expuestos denotan un aumento exponencial del rendimiento en las primeras edades, hasta llegar a los 18 años; los mejores resultados se alcanzan de los 19 a los 23 años y se mantienen hasta los 30. A partir de los 33 años, la curva describe una involución hasta el final del ciclo, con un punto de inflexión más acentuado a los 53 años.

  20. 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)

  1. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    Dupraz, B.; Bertel, E.

    2003-01-01

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

  2. 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.

  3. 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

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

    International Nuclear Information System (INIS)

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

    2013-03-01

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

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

    International Nuclear Information System (INIS)

    Karriem, Z.; Zamonsky, O.M.

    2014-01-01

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

  6. Deposition of RuO{sub 4} on various surfaces in a nuclear reactor containment

    Energy Technology Data Exchange (ETDEWEB)

    Holm, Joachim, E-mail: joachim.holm@chalmers.s [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden); Glaenneskog, Henrik [Ringhals AB, SE-430 22, Vaeroebacka (Sweden); Ekberg, Christian [Department of Nuclear Chemistry, Chalmers University of Technology, Se-412 96 Gothenburg (Sweden)

    2009-07-01

    During a severe nuclear reactor accident with air ingress, ruthenium can be released from the nuclear fuel in the form of ruthenium tetroxide. Hence, it is important to investigate how the reactor containment is able to reduce the source term of ruthenium. The aim of this work was to investigate the deposition of gaseous ruthenium tetroxide on aluminium, copper and zinc, which all appear in relatively large amounts in reactor containment. The experiments show that ruthenium tetroxide is deposited on all the metal surfaces, especially on the copper and zinc surfaces. A large deposition of ruthenium tetroxide also appeared on the relatively inert glass surfaces in the experimental set-ups. The analyses of the different surfaces, with several analytical methods, showed that the form of deposited ruthenium was mainly ruthenium dioxide.

  7. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-02-15

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Subagjo

    2014-07-01

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

  10. PRE design of a molten salt thorium reactor loop

    International Nuclear Information System (INIS)

    Caire, Jean-Pierre; Roure, Anthony

    2007-01-01

    This study is a contribution to the 2004 PCR-RSF program of the Centre National de la Recherche Scientifique (CNRS) devoted to research on high temperature thorium molten salt reactors. A major issue of high temperature molten salt reactors is the very large heat duty to be transferred from primary to secondary loop of the reactor with minimal thermal losses. A possible inner loop made of a series of conventional graphite filter plate exchangers, pipes and pumps was investigated. The loop was assumed to use two counter current flows of the same LiF, BeF 2 , ZrF 4 , UF 4 molten salt flowing through the reactor. The 3D model used the coupling of k-ε turbulent Navier-Stokes equations and thermal applications of the Heat Transfer module of COMSOL Multiphysics. For a reactor delivering 2700 MWth, the model required a set of 114 identical exchangers. Each one was optimized to limit the heat losses to 2882 W. The pipes made of a succession of graphite, ceramics, Hastelloy-N alloy and insulating Microtherm layers led to a thermal loss limited to 550 W per linear meter. In such conditions, the global thermal losses represent only 0.013% of the reactor thermal power for elements covered with an insulator only 3 cm thick. (author)

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

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2012-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

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

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

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

    International Nuclear Information System (INIS)

    1984-11-01

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

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

    International Nuclear Information System (INIS)

    1984-04-01

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

  18. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1958-07-01

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

  20. Construction of 12 EST libraries and characterization of a 12,226 EST dataset for chicory (Cichorium intybus root, leaves and nodules in the context of carbohydrate metabolism investigation

    Directory of Open Access Journals (Sweden)

    Boutry Marc

    2009-01-01

    Full Text Available Abstract Background The industrial chicory, Cichorium intybus, is a member of the Asteraceae family that accumulates fructan of the inulin type in its root. Inulin is a low calories sweetener, a texture agent and a health promoting ingredient due to its prebiotic properties. Average inulin chain length is a critical parameter that is genotype and temperature dependent. In the context of the study of carbohydrate metabolism and to get insight into the transcriptome of chicory root and to visualize temporal changes of gene expression during the growing season, we obtained and characterized 10 cDNA libraries from chicory roots regularly sampled in field during a growing season. A leaf and a nodule libraries were also obtained for comparison. Results Approximately 1,000 Expressed Sequence Tags (EST were obtained from each of twelve cDNA libraries resulting in a 12,226 EST dataset. Clustering of these ESTs returned 1,922 contigs and 4,869 singlets for a total of 6,791 putative unigenes. All ESTs were compared to public sequence databases and functionally classified. Data were specifically searched for sequences related to carbohydrate metabolism. Season wide evolution of functional classes was evaluated by comparing libraries at the level of functional categories and unigenes distribution. Conclusion This chicory EST dataset provides a season wide outlook of the genes expressed in the root and to a minor extent in leaves and nodules. The dataset contains more than 200 sequences related to carbohydrate metabolism and 3,500 new ESTs when compared to other recently released chicory EST datasets, probably because of the season wide coverage of the root samples. We believe that these sequences will contribute to accelerate research and breeding of the industrial chicory as well as of closely related species.

  1. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

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

  2. Comparison of MCNPX-C90 and TRIPOLI-4-D for fuel depletion calculations of a Gas-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Reyes-Ramirez, Ricardo; Martin-del-Campo, Cecilia; Francois, Juan-Luis; Brun, Emeric; Dumonteil, Eric; Malvagi, Fausto

    2010-01-01

    The Gas-cooled Fast Reactor is one of the reactor concepts selected by the Generation IV International Forum for the next generation of innovative nuclear energy systems. Several fuel design concepts are being investigated. Burnup depletion of mixed fuel of uranium and plutonium, cooled with gas in a fast neutron energy spectrum must be simulated. Various codes are being developed and/or adapted to improve the quality of the results, and also to reduce the computing time required for the simulations. The main objective of this work is to compare the fuel depletion results obtained with MCNPX-CINDER90 code and the new TRIPOLI-4-Depletion code (developed by the Commissariat a l'Energie Atomique) of a fuel design concept for the Gas-cooled Fast Reactor. Calculations were made for an equivalent homogeneous model of fuel rods in a hexagonal mesh assembly. Total reflection conditions were applied on the six lateral faces and the two axial faces of the assembly. The materials used in the fuel assembly are: carbide of uranium and plutonium as fuel, silicon carbide as cladding, and helium gas as coolant. JEFF libraries of effective cross sections were used in both codes. Two methods of burnup step calculations were performed with TRIPOLI-4-D, the Euler and the CSADA, and their results were compared with the MCNPX-CINDER90 CSADA method. A period of 300 days of irradiation time was considered, which was divided into 12 steps. Results of the infinite multiplication factor as function of the irradiation time, and the evolution of the isotope concentrations for a selected group of nuclides were compared. The main conclusion is that very similar results were obtained for the three types of depletion calculations which were compared: (1) MCNPX-C90 CSADA; (2) TRIPOLI-4-D CSADA, and (3) TRIPOLI-4-D EULER. The best calculation time was obtained with the TRIPOLI-4-D EULER method, which needed approximately half the time than the other two. In summary, it is sufficiently good to use

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Van Heek, A.I.

    1996-06-01

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

  5. G2 and G3 reactors design; Description des reacteurs G2 et G3

    Energy Technology Data Exchange (ETDEWEB)

    Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

    1958-07-01

    'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

    Guentay, S.

    1996-12-01

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

  9. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  10. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  13. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  14. Feasibility study of a magnetic fusion production reactor

    Science.gov (United States)

    Moir, R. W.

    1986-12-01

    A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  16. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Mirvakili, S.M.

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  17. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  18. Biodegradation of 4-bromophenol by Arthrobacter chlorophenolicus A6T in a newly designed packed bed reactor.

    Science.gov (United States)

    Sahoo, Naresh Kumar; Ghosh, Pranab Kumar; Pakshirajan, Kannan

    2013-02-01

    Bromophenol is listed as a priority pollutant by the U.S. EPA. However, there has been no report on the removal of bromophenol in any biological system that is operated in a continuous mode. The efficiency of Arthrobacter chlorophenolicus A6(T) on the biodegradation of 4-bromophenol (4-BP) in a newly designed packed bed reactor (PBR) was evaluated with different influent 4-BP concentrations between 400 mg l(-1) and 1200 mg l(-1) and hydraulic retention times (HRTs) between 24 h and 7.5 h. The response of the PBR to 4-BP shock loadings was also tested, and the bioreactor was found to adequately handle these shock loadings. The percentage of effluent toxicity in the PBR was tested using mixed microbial consortia as the test species; this experiment was performed using a 4-BP influent concentration of 1200 mg l(-1) and HRTs between 24 h and 7.5 h. A maximal 98% effluent toxicity removal was achieved when the PBR was operated at an HRT of 24 h. In the present study, 4-BP was used as the sole source of carbon and energy, and the complete removal of 4-BP was achieved with 4-BP loading rates of up to 2277 mg l(-1) day(-1). Copyright © 2012 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  19. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Science.gov (United States)

    Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

    2015-12-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

  20. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    International Nuclear Information System (INIS)

    Hamann, S.; Röpcke, J.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.

    2015-01-01

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH 4 , C 2 H 2 , HCN, and NH 3 ). With the help of OES, the rotational temperature of the screen plasma could be determined

  1. Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

    Energy Technology Data Exchange (ETDEWEB)

    Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J. [INP-Greifswald, Felix-Hausdorff-Str. 2, 17489 Greifswald (Germany); Börner, K.; Burlacov, I.; Spies, H.-J. [TU Bergakademie Freiberg, Institute of Materials Engineering, Gustav-Zeuner-Str. 5, 09599 Freiberg (Germany); Strämke, M.; Strämke, S. [ELTRO GmbH, Arnold-Sommerfeld-Ring 3, 52499 Baesweiler (Germany)

    2015-12-15

    A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.

  2. 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.)

  3. A new degassing membrane coupled upflow anaerobic sludge blanket (UASB) reactor to achieve in-situ biogas upgrading and recovery of dissolved CH4 from the anaerobic effluent

    DEFF Research Database (Denmark)

    Luo, Gang; Wang, Wen; Angelidaki, Irini

    2014-01-01

    A new technology for in-situ biogas upgrading and recovery of CH4 from the effluent of biogas reactors was proposed and demonstrated in this study. A vacuum degassing membrane module was used to desorb CO2 from the liquid phase of a biogas reactor. The degassing membrane was submerged...... into a degassing unit (DU). The results from batch experiments showed that mixing intensity, transmembrane pressure, pH and inorganic carbon concentration affected the CO2 desorption rate in the DU. Then, the DU was directly connected to an upflow anaerobic sludge blanket (UASB) reactor. The results showed the CH4...... content was only 51.7% without desorption of CO2, while it increased when the liquid of UASB was recycled through the DU. The CH4 content increased to 71.6%, 90%, and 94% with liquid recirculation rate through the DU of 0.21, 0.42 and 0.63L/h, respectively. The loss of methane due to dissolution...

  4. Purification, crystallization and preliminary crystallographic analysis of Est-Y29: a novel oligomeric β-lactamase

    International Nuclear Information System (INIS)

    Kim, SeungBum; Joo, Sangbum; Yoon, Sangyoung; Kim, Sungsoo; Moon, Jongkook; Ryu, Yeonwoo; Kim, Kyeong Kyu; Kim, T. Doohun

    2009-01-01

    Est-Y29, a novel oligomeric class C β-lactamase from a metagenomic library, was crystallized in space group I4 1 and diffraction data were collected to 1.49 Å resolution. β-Lactam antibiotics such as penicillins and cephalosporins have a four-atom ring as a common element in their structure. The β-lactamases, which catalyze the inactivation of these antibiotics, are of great interest because of their high incidence in pathogenic bacteria. A novel oligomeric class C β-lactamase (Est-Y29) from a metagenomic library was expressed, purified and crystallized. The recombinant protein was expressed in Escherichia coli with an N-terminal 6×His tag and purified to homogeneity. EstY-29 was crystallized and X-ray intensity data were collected to 1.49 Å resolution using synchrotron radiation

  5. Solution of a benchmark set problems for BWR and PWR reactors with UO2 and MOX fuels using CASMO-4

    International Nuclear Information System (INIS)

    Martinez F, M.A.; Valle G, E. del; Alonso V, G.

    2007-01-01

    In this work some of the results for a group of benchmark problems of light water reactors that allow to study the physics of the fuels of these reactors are presented. These benchmark problems were proposed by Akio Yamamoto and collaborators in 2002 and they include two fuel types; uranium dioxide (UO 2 ) and mixed oxides (MOX). The range of problems that its cover embraces three different configurations: unitary cell for a fuel bar, fuel assemble of PWR and fuel assemble of BWR what allows to carry out an understanding analysis of the problems related with the fuel performance of new generation in light water reactors with high burnt. Also these benchmark problems help to understand the fuel administration in core of a BWR like of a PWR. The calculations were carried out with CMS (of their initials in English Core Management Software), particularly with CASMO-4 that is a code designed to carry out analysis of fuels burnt of fuel bars cells as well as fuel assemblies as much for PWR as for BWR and that it is part in turn of the CMS code. (Author)

  6. Technical note: Development of a Linear Flow Channel Reactor for ...

    African Journals Online (AJOL)

    Technical note: Development of a Linear Flow Channel Reactor for sulphur removal ... AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... 000 mg∙ℓ-1 Na2SO4 solution) and the Liner Flow Channel Reactors (surface area ...

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

    International Nuclear Information System (INIS)

    Yang Fanglin; Yang Yingli; Wang Guangshen

    1989-10-01

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

  9. The molten salt reactor adventure

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1985-01-01

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

  10. 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

  11. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

  12. A experiência estética sob um olhar fenomenológico

    Directory of Open Access Journals (Sweden)

    Alice Casanova Reis

    2011-01-01

    Full Text Available Este artículo describe la experiencia estética como un fenómeno centrado en la percepción sensorial. El trabajo se desarrolla a partir de una investigación de la psicología social basado en la fenomenología, sobre todo en los planteamientos de Merleau-Ponty y Dufrenne. Concibe la experiencia estética como un modo de relación mediada por la percepción sensorial de un objeto estético. Se analizan diversos aspectos de la experiencia estética: la dimensión relacional y sensible; ¿cómo se constituye el sujeto y el objeto estético; la dimensión corporal de la experiencia estética; la intencionalidad en la estética; la experiencia estética de la naturaleza. Llegamos a la conclusión de que hay un significado ontológico de la experiencia estética, ya que proporciona una apertura a la diferencia por el contacto con la alteridad, y una nueva mirada a la realidad. Se destaca al final la relevancia de esta experiencia a la Psicología Social.

  13. Development of EST-derived markers in Dendrobium from EST of related taxa

    Directory of Open Access Journals (Sweden)

    Narisa Juejun

    2013-04-01

    Full Text Available Public databases are useful for molecular marker development. The major aim of this study was to develop expressedsequence tag (EST-derived markers in Dendrobium from available ESTs of Phalaenopsis and Dendrobium. A total of 6063sequences were screened for simple sequence repeats (SSRs and introns. Primers flanking these regions were generated andtested on genomic DNAs of Phalaenopsis and Dendrobium. Twenty-three percent of amplifiable Phalaenopsis EST-derivedmarkers were cross-genera transferable to Dendrobium. Forty-one markers from both Phalaenopsis and Dendrobium thatamplified in Dendrobium were assessed on six commercial cultivars and six wild accessions. All of them were transferableamong Dendrobium species. High polymorphism and heterozygosity were observed within wild accessions. Sixteen polymorphic markers were evaluated for linkage analysis on an F1 segregating population. Seven markers were mapped into threelinkage groups, two of which showed syntenic relationship between dendrobium and rice. This relationship will facilitatefurther quantitative trait loci (QTL mapping and comparative genomic studies of Dendrobium. Our results indicate thatPhalaenopsis EST-derived markers are valuable tools for genetic research and breeding applications in Dendrobium.

  14. Mechanical design of a light water breeder reactor

    International Nuclear Information System (INIS)

    Fauth, W.L. Jr.; Jones, D.S.; Kolsun, G.J.; Erbes, J.G.; Brennan, J.J.; Weissburg, J.A.; Sharbaugh, J.E.

    1976-01-01

    In a light water reactor system using the thorium-232--uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements. 4 claims, 24 drawing figures

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. A new degassing membrane coupled upflow anaerobic sludge blanket (UASB) reactor to achieve in-situ biogas upgrading and recovery of dissolved CH4 from the anaerobic effluent

    International Nuclear Information System (INIS)

    Luo, Gang; Wang, Wen; Angelidaki, Irini

    2014-01-01

    Highlights: • A new UASB configuration was developed by coupling with degassing membrane. • In-situ biogas upgrading was achieved with high methane content (>90%). • Decrease of dissolved methane in the anaerobic effluent was achieved. - Abstract: A new technology for in-situ biogas upgrading and recovery of CH 4 from the effluent of biogas reactors was proposed and demonstrated in this study. A vacuum degassing membrane module was used to desorb CO 2 from the liquid phase of a biogas reactor. The degassing membrane was submerged into a degassing unit (DU). The results from batch experiments showed that mixing intensity, transmembrane pressure, pH and inorganic carbon concentration affected the CO 2 desorption rate in the DU. Then, the DU was directly connected to an upflow anaerobic sludge blanket (UASB) reactor. The results showed the CH 4 content was only 51.7% without desorption of CO 2 , while it increased when the liquid of UASB was recycled through the DU. The CH 4 content increased to 71.6%, 90%, and 94% with liquid recirculation rate through the DU of 0.21, 0.42 and 0.63 L/h, respectively. The loss of methane due to dissolution in the effluent was reduced by directly pumping the reactor effluent through the DU. In this way, the dissolved CH 4 concentration in the effluent decreased from higher than 0.94 mM to around 0.13 mM, and thus efficient recovery of CH 4 from the anaerobic effluent was achieved. In the whole operational period, the COD removal efficiency and CH 4 yield were not obviously affected by the gas desorption

  17. Probabilistic safety criteria on high burnup HWR fuels

    International Nuclear Information System (INIS)

    Marino, A.C.

    2002-01-01

    BACO is a code for the simulation of the thermo-mechanical and fission gas behaviour of a cylindrical fuel rod under operation conditions. Their input parameters and, therefore, output ones may include statistical dispersion. In this paper, experimental CANDU fuel rods irradiated at the NRX reactor together with experimental MOX fuel rods and the IAEA-CRP FUMEX cases are used in order to determine the sensitivity of BACO code predictions. The techniques for sensitivity analysis defined in BACO are: the 'extreme case analysis', the 'parametric analysis' and the 'probabilistic (or statistics) analysis'. We analyse the CARA and CAREM fuel rods relation between predicted performance and statistical dispersion in order of enhanced their original designs taking account probabilistic safety criteria and using the BACO's sensitivity analysis. (author)

  18. Neurexin Dysfunction in Adult Neurons Results in Autistic-like Behavior in Mice

    Directory of Open Access Journals (Sweden)

    Luis G. Rabaneda

    2014-07-01

    Full Text Available Autism spectrum disorders (ASDs comprise a group of clinical phenotypes characterized by repetitive behavior and social and communication deficits. Autism is generally viewed as a neurodevelopmental disorder where insults during embryonic or early postnatal periods result in aberrant wiring and function of neuronal circuits. Neurexins are synaptic proteins associated with autism. Here, we generated transgenic βNrx1ΔC mice in which neurexin function is selectively impaired during late postnatal stages. Whole-cell recordings in cortical neurons show an impairment of glutamatergic synaptic transmission in the βNrx1ΔC mice. Importantly, mutant mice exhibit autism-related symptoms, such as increased self-grooming, deficits in social interactions, and altered interaction for nonsocial olfactory cues. The autistic-like phenotype of βNrx1ΔC mice can be reversed after removing the mutant protein in aged animals. The defects resulting from disruption of neurexin function after the completion of embryonic and early postnatal development suggest that functional impairment of mature circuits can trigger autism-related phenotypes.

  19. 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.

  20. Enhanced nitrogen removal from piggery wastewater with high NH4+ and low COD/TN ratio in a novel upflow microaerobic biofilm reactor.

    Science.gov (United States)

    Meng, Jia; Li, Jiuling; Li, Jianzheng; Antwi, Philip; Deng, Kaiwen; Nan, Jun; Xu, Pianpian

    2018-02-01

    To enhance nutrient removal more cost-efficiently in microaerobic process treating piggery wastewater characterized by high ammonium (NH 4 + -N) and low chemical oxygen demand (COD) to total nitrogen (TN) ratio, a novel upflow microaerobic biofilm reactor (UMBR) was constructed and the efficiency in nutrient removal was evaluated with various influent COD/TN ratios and reflux ratios. The results showed that the biofilm on the carriers had increased the biomass in the UMBR and enhanced the enrichment of slow-growth-rate bacteria such as nitrifiers, denitrifiers and anammox bacteria. The packed bed allowed the microaerobic biofilm process perform well at a low reflux ratio of 35 with a NH 4 + -N and TN removal as high as 93.1% and 89.9%, respectively. Compared with the previously developed upflow microaerobic sludge reactor, the UMBR had not changed the dominant anammox approach to nitrogen removal, but was more cost-efficiently in treating organic wastewater with high NH 4 + -N and low COD/TN ratio. Copyright © 2017 Elsevier Ltd. All rights reserved.

  1. 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

  2. Elements on reactor control

    International Nuclear Information System (INIS)

    Bruna, G.B.

    1998-01-01

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

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

    Directory of Open Access Journals (Sweden)

    César González-Ugalde

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-01

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

  5. ConiferEST: an integrated bioinformatics system for data reprocessing and mining of conifer expressed sequence tags (ESTs).

    Science.gov (United States)

    Liang, Chun; Wang, Gang; Liu, Lin; Ji, Guoli; Fang, Lin; Liu, Yuansheng; Carter, Kikia; Webb, Jason S; Dean, Jeffrey F D

    2007-05-29

    With the advent of low-cost, high-throughput sequencing, the amount of public domain Expressed Sequence Tag (EST) sequence data available for both model and non-model organism is growing exponentially. While these data are widely used for characterizing various genomes, they also present a serious challenge for data quality control and validation due to their inherent deficiencies, particularly for species without genome sequences. ConiferEST is an integrated system for data reprocessing, visualization and mining of conifer ESTs. In its current release, Build 1.0, it houses 172,229 loblolly pine EST sequence reads, which were obtained from reprocessing raw DNA sequencer traces using our software--WebTraceMiner. The trace files were downloaded from NCBI Trace Archive. ConiferEST provides biologists unique, easy-to-use data visualization and mining tools for a variety of putative sequence features including cloning vector segments, adapter sequences, restriction endonuclease recognition sites, polyA and polyT runs, and their corresponding Phred quality values. Based on these putative features, verified sequence features such as 3' and/or 5' termini of cDNA inserts in either sense or non-sense strand have been identified in-silico. Interestingly, only 30.03% of the designated 3' ESTs were found to have an authenticated 5' terminus in the non-sense strand (i.e., polyT tails), while fewer than 5.34% of the designated 5' ESTs had a verified 5' terminus in the sense strand. Such previously ignored features provide valuable insight for data quality control and validation of error-prone ESTs, as well as the ability to identify novel functional motifs embedded in large EST datasets. We found that "double-termini adapters" were effective indicators of potential EST chimeras. For all sequences with in-silico verified termini/terminus, we used InterProScan to assign protein domain signatures, results of which are available for in-depth exploration using our biologist

  6. ConiferEST: an integrated bioinformatics system for data reprocessing and mining of conifer expressed sequence tags (ESTs

    Directory of Open Access Journals (Sweden)

    Carter Kikia

    2007-05-01

    Full Text Available Abstract Background With the advent of low-cost, high-throughput sequencing, the amount of public domain Expressed Sequence Tag (EST sequence data available for both model and non-model organism is growing exponentially. While these data are widely used for characterizing various genomes, they also present a serious challenge for data quality control and validation due to their inherent deficiencies, particularly for species without genome sequences. Description ConiferEST is an integrated system for data reprocessing, visualization and mining of conifer ESTs. In its current release, Build 1.0, it houses 172,229 loblolly pine EST sequence reads, which were obtained from reprocessing raw DNA sequencer traces using our software – WebTraceMiner. The trace files were downloaded from NCBI Trace Archive. ConiferEST provides biologists unique, easy-to-use data visualization and mining tools for a variety of putative sequence features including cloning vector segments, adapter sequences, restriction endonuclease recognition sites, polyA and polyT runs, and their corresponding Phred quality values. Based on these putative features, verified sequence features such as 3' and/or 5' termini of cDNA inserts in either sense or non-sense strand have been identified in-silico. Interestingly, only 30.03% of the designated 3' ESTs were found to have an authenticated 5' terminus in the non-sense strand (i.e., polyT tails, while fewer than 5.34% of the designated 5' ESTs had a verified 5' terminus in the sense strand. Such previously ignored features provide valuable insight for data quality control and validation of error-prone ESTs, as well as the ability to identify novel functional motifs embedded in large EST datasets. We found that "double-termini adapters" were effective indicators of potential EST chimeras. For all sequences with in-silico verified termini/terminus, we used InterProScan to assign protein domain signatures, results of which are available

  7. Butanol production by Clostridium acetobutylicum in a continuous packed bed reactor.

    Science.gov (United States)

    Napoli, Fabio; Olivieri, Giuseppe; Russo, Maria Elena; Marzocchella, Antonio; Salatino, Piero

    2010-06-01

    In this study, we report on a butanol production process by immobilized Clostridium acetobutylicum in a continuous packed bed reactor (PBR) using Tygon rings as a carrier. The medium was a solution of lactose (15-30 g/L) and yeast extract (3 g/L) to emulate the cheese whey, an abundant lactose-rich wastewater. The reactor was operated under controlled conditions with respect to the pH and to the dilution rate. The pH and the dilution rate ranged between 4 and 5, the dilution rate between 0.54 and 2.4 h(-1) (2.5 times the maximum specific growth rate assessed for suspended cells). The optimal performance of the reactor was recorded at a dilution rate of 0.97 h(-1): the butanol productivity was 4.4 g/Lh and the selectivity of solvent in butanol was 88%(w).

  8. La digestión anaerobia y los reactores UASB. Generalidades

    OpenAIRE

    Yaniris Lorenzo; Ma. Cristina Obaya

    2006-01-01

    Se muestran las generalidades de los reactores, se da a conocer su concepto, se enumeran las aguas residuales que pueden ser tratadas en los mismos, se comentan los parámetros a tener en cuenta para que funcione adecuadamente y se enumeran las ventajas y desventajas de este proceso, así como su aplicabilidad.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. Measurements of low reactivities using a reactor oscillator

    International Nuclear Information System (INIS)

    Obradovic, D.; Petrovic, M.

    1965-12-01

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  11. Measurements of low reactivities using a reactor oscillator

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D; Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-12-15

    Most of the methods of measuring reactivity are limited to the region from several hundreds to several thousands of pcm. The present work develops a method of measuring low reactivities from several pcm to about 600 pcm using the ROB-1 reactor oscillator on the RB reactor of the Boris Kidric Institute of Nuclear Sciences at Vinca. The accuracy of measurement is better than 1%. Several methods are used to measure low reactivities. The most often used is the method based on measuring the stable reactor period. The bottom limit of this method is about 30 porn /1,2/. For control rod calibration the method of rod oscillation is used /3,4/. This method is confronted with considerable influence of space effects /5/. Reference /6/ reports on a method for measuring the reactivity coefficient at a critical level in liquid-moderated reactors. The method is based on measuring reactor response to the oscillation of the moderator about the critical level. The present work reports on a method of determining the reactivity by measuring the phase shift between the perturbation of the effective multiplication factor and reactor response. With the use of the ROB-1 reactor oscillator, the method allows measurement of the reactivity from several pcm to about 600 pcm with an accuracy of 1% (author)

  12. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

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

  13. Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

    International Nuclear Information System (INIS)

    Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

    2010-01-01

    SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

  14. Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

    International Nuclear Information System (INIS)

    Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

    2013-01-01

    A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO

  15. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

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

    1974-07-01

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

  16. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

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

    1974-01-01

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

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

    International Nuclear Information System (INIS)

    Bowman, S.M.; Suto, T.

    1996-10-01

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

  18. 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

  19. Sources of gamma radiation in a reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-05-15

    In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

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

    International Nuclear Information System (INIS)

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

    2014-03-01

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

  1. The IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    2012-12-01

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

  3. Prime note sulla tutela penale dei culti nei Paesi dell’Est Europa

    Directory of Open Access Journals (Sweden)

    Giovanni Cimbalo

    2011-05-01

    Full Text Available Testo della relazione tenuta al Convegno “La Carta e la Corte” (Ferrara, 27 ottobre 2007 destinata alla pubblicazione negli Atti. SOMMARIO: 1. Alcune considerazioni preliminari sullo status delle Confessioni religiose nei paesi dell’Est Europa - 2. I nuovi orientamenti del diritto penale nell’Est Europa - 3. Le norme statali in materia di tutela penale dei culti e del sentimento religioso. 3. Le norme penali relative ai culti e a al sentimento religioso prima del 1992 nei Paesi dell’Est Europa - 4. Tipologie e tecniche legislative di tutela penale dei culti dopo il 1992 nei Paesi dell’Est Europa - 5. Alcune sommarie considerazioni.

  4. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  5. Predicted sub-populations in a marine shrimp proteome as revealed by combined EST and cDNA data from multiple Penaeus species

    Directory of Open Access Journals (Sweden)

    Kotewong Rattanawadee

    2010-11-01

    Full Text Available Abstract Background Many species of marine shrimp in the Family Penaeidae, viz. Penaeus (Litopenaeus vannamei, Penaeus monodon, Penaeus (Fenneropenaeus chinensis, and Penaeus (Marsupenaeus japonicus, are animals of economic importance in the aquaculture industry. Yet information about their DNA and protein sequences is lacking. In order to predict their collective proteome, we combined over 270,000 available EST and cDNA sequences from the 4 shrimp species with all protein sequences of Drosophila melanogaster and Caenorhabditis elegans. EST data from 4 other crustaceans, the crab Carcinus maenas, the lobster Homarus americanus (Decapoda, the water flea Daphnia pulex, and the brine shrimp Artemia franciscana were also used. Findings Similarity searches from EST collections of the 4 shrimp species matched 64% of the protein sequences of the fruit fly, but only 45% of nematode proteins, indicating that the shrimp proteome content is more similar to that of an insect than a nematode. Combined results with 4 additional non-shrimp crustaceans increased matching to 78% of fruit fly and 56% of nematode proteins, suggesting that present shrimp EST collections still lack sequences for many conserved crustacean proteins. Analysis of matching data revealed the presence of 4 EST groups from shrimp, namely sequences for proteins that are both fruit fly-like and nematode-like, fruit fly-like only, nematode-like only, and non-matching. Gene ontology profiles of proteins for the 3 matching EST groups were analyzed. For non-matching ESTs, a small fraction matched protein sequences from other species in the UniProt database, including other crustacean-specific proteins. Conclusions Shrimp ESTs indicated that the shrimp proteome is comprised of sub-populations of proteins similar to those common to both insect and nematode models, those present specifically in either model, or neither. Combining small EST collections from related species to compensate for their

  6. Safety aspects on dependability management for a TRIGA research reactor in Romania

    International Nuclear Information System (INIS)

    Vieru, Gheorghe

    2002-01-01

    Safety on the management for a nuclear research reactor involves a 'good dependability management' of the activities, such as: reliability, availability, maintainability and maintenance support. In order to evaluate the safety management aspects intended to be applied at a research reactor management, the performance dependability indicators and their impact over the availability and reactor safety have to be established. The document ISO 9000-4/IEC 300-1 'Dependability Management' (1995), describes five internationally agreed indicators of the reactor equipment dependability, each of them can be used for corrective maintenance or for preventive maintenance, such as: I 1 - equipment Maintenance Frequency; I 2 - equipment Maintenance Effort; I-3 - equipment Maintenance Downtime Factor; I 4 - equipment Maintenance Contribution to the System Function Downtime Factor; I 5 - equipment Maintenance Contribution to the reactor Capability Loss Factor. The paper presents an evaluation of those 5 mentioned indicators with referring only at the primary circuit of the INR's TRIGA research reactor and conclusion. The analyzed period was stated between 1994-1999. It is to be noted that this type of analyze is performed for the first time for a research reactor. (author)

  7. Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Sembiring, Tagor Malem

    2012-01-01

    Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

  8. Tratamento anaeróbio de águas residuárias do beneficiamento de café por via úmida em reatores UASB em dois estágios Anaerobic treatment of wastewater from coffee pulping in upflow anaerobic sludge blanquet (UASB in two stages

    Directory of Open Access Journals (Sweden)

    Marcelo Bruno

    2008-06-01

    Full Text Available Neste trabalho, avaliou-se a eficiência do tratamento de águas residuárias do beneficiamento de café por via úmida em reatores anaeróbios de fluxo ascendente com manta de lodo (UASB, em dois estágios, em escala de bancada, submetidos a tempos de detenção hidráulica (TDH de 4,0; 5,2 e 6,2 dias e cargas orgânicas volumétricas (COV de 5,8; 3,0 e 3,6 g DQO total (L d-1, no primeiro reator (R1, e TDH de 2,0; 2,6 e 3,1 dias e COV de 5,8; 0,5 e 0,4 g DQO total (L d-1, no segundo reator (R2. Os valores médios de DQO do afluente variaram de 15.440 a 23.040 mg O2 L-1. As eficiências médias de remoção de DQO total e SST foram de 66 a 98% e de 93 a 97%, respectivamente, nos reatores UASB, em dois estágios. O teor médio de metano no biogás variou de 69 a 89%, no reator R1, e de 52 a 73%, no reator R2. A produção volumétrica máxima de metano de 0,708 L CH4 (L reator d-1 foi obtida com COV de 3,6 g DQO (Ld-1 e TDH de 6,2 d, no reator R1. Os valores médios de pH variaram de 4,7 a 7,7 e de 4,9 a 8,0 nos efluentes dos reatores R1 e R2, respectivamente. As concentrações de ácidos voláteis totais nos efluentes mantiveram-se estáveis com valores inferiores a 100 mg L-1, com TDH de 5,2 e 6,2 dias, no reator R1, e TDH de 2,6 e 3,1 dias, no reator R2. As concentrações médias de fenóis totais no afluente variaram de 80 a 97 mg L-1 e as eficiências médias de remoção nos reatores UASB, em dois estágios, foram de 72 a 90%.In this work, it was evaluated the efficiency of two stage up flow anaerobic sludge blanquet (UASB reactors, in bench scale, treating a liquid effluent from the coffee pulping, submitted to hydraulic detention time (HDT of 4.0; 5.2 and 6.2 days, resulting in organic loading rate (OLR of 5.8; 3.6 and 3.0 g COD (L d-1 in the first reactor (R1 and HDT of 2.0; 2.6 and 3.1 days with OLR of 5.8; 0.5 and 0.4 g DQO total (L d-1 in the second reactor (R2. The medium values of total COD affluent varied from 15,440 to 23,040 mg

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-04-01

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

  10. Identification of candidate genes for human pituitary development by EST analysis

    Directory of Open Access Journals (Sweden)

    Xiao Huasheng

    2009-03-01

    Full Text Available Abstract Background The pituitary is a critical neuroendocrine gland that is comprised of five hormone-secreting cell types, which develops in tandem during the embryonic stage. Some essential genes have been identified in the early stage of adenohypophysial development, such as PITX1, FGF8, BMP4 and SF-1. However, it is likely that a large number of signaling molecules and transcription factors essential for determination and terminal differentiation of specific cell types remain unidentified. High-throughput methods such as microarray analysis may facilitate the measurement of gene transcriptional levels, while Expressed sequence tag (EST sequencing, an efficient method for gene discovery and expression level analysis, may no-redundantly help to understand gene expression patterns during development. Results A total of 9,271 ESTs were generated from both fetal and adult pituitaries, and assigned into 961 gene/EST clusters in fetal and 2,747 in adult pituitary by homology analysis. The transcription maps derived from these data indicated that developmentally relevant genes, such as Sox4, ST13 and ZNF185, were dominant in the cDNA library of fetal pituitary, while hormones and hormone-associated genes, such as GH1, GH2, POMC, LHβ, CHGA and CHGB, were dominant in adult pituitary. Furthermore, by using RT-PCR and in situ hybridization, Sox4 was found to be one of the main transcription factors expressed in fetal pituitary for the first time. It was expressed at least at E12.5, but decreased after E17.5. In addition, 40 novel ESTs were identified specifically in this tissue. Conclusion The significant changes in gene expression in both tissues suggest a distinct and dynamic switch between embryonic and adult pituitaries. All these data along with Sox4 should be confirmed to further understand the community of multiple signaling pathways that act as a cooperative network that regulates maturation of the pituitary. It was also suggested that EST

  11. Ortodoncia estética invisible

    OpenAIRE

    Chávez Sevillano, Manuel Gustavo; Soldevilla Galarza, Luciano

    2014-01-01

    El adulto suele ofrecer resistencia a los convencionales tratamientos ortodóncicos, debido a la necesidad de llevar brackets visibles, tanto metálicos como estéticos de porcelana. El concepto de Ortodoncia Estética Invisible u Ortodoncia Lingual cubre las expectativas de este tipo de pacientes. La técnica multibrackets con aparatología lingual tiene aproximadamente 25 años de desarrollo y con la experiencia de los casos tratados, se ha llegado a la concepción de una técnica completamente prot...

  12. Studies of conceptual spheromak fusion reactors

    International Nuclear Information System (INIS)

    Katsurai, M.; Yamada, M.

    1982-01-01

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

  13. Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

    International Nuclear Information System (INIS)

    Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

    1975-01-01

    A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

  14. 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)

  15. 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.

  16. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

  17. Development of a coupled reactor with a catalytic combustor and steam reformer for a 5 kW solid oxide fuel cell system

    International Nuclear Information System (INIS)

    Kang, Sanggyu; Lee, Kanghun; Yu, Sangseok; Lee, Sang Min; Ahn, Kook-Young

    2014-01-01

    Highlights: • Proposes the scale-up strategy to develop a large-scale coupled reactor. • Investigation of performance of steam reformer coupled with catalytic combustor. • Experimental parameters are inlet temp., air excess ratio, SCR, fuel utilization. • Evaluation of the heat transfer distribution along the gas flow direction. • The mean value of methane conversion rate is approximately 93.4%. - Abstract: The methane (CH 4 ) conversion rate of a steam reformer can be increased by thermal integration with a catalytic combustor, called a coupled reactor. In the present study, a 5 kW coupled reactor has been developed based on a 1 kW coupled reactor in previous work. The geometric parameters of the space velocity, diameter and length of the coupled reactor selected from the 1 kW coupled reactor are tuned and applied to the design of the 5 kW coupled reactor. To confirm the scale-up strategy, the performance of 5 kW coupled reactor is experimentally investigated with variations of operating parameters such as the fuel utilization in the solid oxide fuel cell (SOFC) stack, the inlet temperature of the catalytic combustor, the excess air ratio of the catalytic combustor, and the steam to carbon ratio (SCR) in the steam reformer. The temperature distributions of coupled reactors are measured along the gas flow direction. The gas composition at the steam reformer outlet is measured to find the CH 4 conversion rate of the coupled reactor. The maximum value of the CH 4 conversion rate is approximately 93.4%, which means the proposed scale-up strategy can be utilized to develop a large-scale coupled reactor

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  19. Results of a comparison study of advanced reactors

    International Nuclear Information System (INIS)

    Bueno de Mesquita, K.G.; Gout, W.; Heil, J.A.; Tanke, R.H.J.; Geevers, F.

    1991-06-01

    The PINK programme is a 4-year programme of five parties involved in nuclear energy in the Netherlands: GKN (operator of the Dodewaard plant), KEMA (Research institute of the Netherlands Utilities), ECN (Netherlands Energy Research Foundation), NUCON (Engineering and Contracting Company) and IRI Interfaculty Reactor Institute of the Delft University of Technology), to coordinate their efforts to intensify the nuclear competence of the industry, the utilities and the research and engineering companies. This programme is sponsored by the Ministry of Economic Affairs. The PINK programme consists of five parts. This report pertains to part 1 of the programme: comparison study of advanced reactors concerning the four so-called second-stage designs SBWR, AP600, SIR and CANDU, which, compared to the first-stage reactor designs, features increased use of passive safety systems and simplification. The objective of the current study is to compare these advanced reactor designs in order to provide comprehensive information for the PINK steering committee that is useful in the selection process of a design for further study and development work. In ch. 2 the main features of the four reactors are highlighted. In ch. 3 the most important safety features and the behaviour of the four reactors under accident situations are compared. Passive safety systems are identified and forgivingness is described and compared. Results of the preliminary probabilistic safety analysis are presented. Ch. 4 deals with the proven technology of the four concepts, ch. 5 with the Netherlands requirements, ch. 6 with commercial aspects, and ch. 7 with the fuel cycle and radioactive waste produced. In ch. 8 the costs are compared and finally in ch. 9 conclusions are drawn and recommendations are made. (author). 13 figs

  20. Problems related with the power regulation of reactors by physico-chemical methods, and the behaviour of water and heavy water in nuclear reactors; Comportement de l'eau et de l'eau lourde dans les reacteurs nucleaires et problemes de la regulation de puissance par voie physico-chimique

    Energy Technology Data Exchange (ETDEWEB)

    Dolle, L; Conan, D; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Experience of the CEA heavy water reactors and a systematic study of the radiolytic decomposition of water in the core of swimming-pool reactors are described. Setting up of reactivity control by physico-chemical methods. Reactivity control by homogeneous poisoning of the reactor A comparison of the evolution of xenon poisoning with the residual anti reactivity of the poison in solution during its nuclear consumption establishes the programme which must govern the variation in its concentration if the exact compensation is to be produced The behaviour of the poison towards the reactor materials under the particular operational conditions must be taken into account. Radiolytic decomposition of water in the reactors in the presence of soluble poisons: A study of the effect of certain chemically inert salts, present in small concentrations in the water, on its radiolytic decomposition rate, has led to some new results which are discussed. The choice of a soluble poison is justified on the basis of the above results. Reactivity control by the use of a gaseous absorbent The use of a gas control rod circuit for compensation purposes, in place of solid control rods is described. The use of soluble poisons in the moderator to compensate the xenon effect, and of a gaseous absorbent in a circuit known as a gas control rod form original aspects of the reactivity control in the reactor EL 4. (authors) [French] L'observation du comportement de l'eau et de l'eau lourde dans les reacteurs en exploitation, contribue au fonctionnement sur de ceux-ci et oriente certaines etudes relatives aux techniques de controle de la reactivite par mise en oeuvre de poisons solubles. L'utilisation de poisons nucleaires dissous dans l'eau du reacteur entraine une pollution chimique de celle-ci. Les conditions d'emploi permettant d'eviter les effets indesirables de cette pollution sont etudiees. Les problemes analytiques - bien qu'importants - ne sont pas abordes dans le cadre de la communication

  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. 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

  3. Cultivares e estádios de Colheita no Rendimento Forrageiro da Soja

    Directory of Open Access Journals (Sweden)

    Pedro Milanez de Rezende

    2011-04-01

    Full Text Available Com o objetivo de verificar a influência do estádio de colheita sobre a produção e a qualidade do feno de duas cultivares de soja foi conduzido ensaio em Lavras, MG, no ano agrícola 2001/02. Utilizou-se o delineamento experimental de blocos casualizados, em esquema fatorial 2 x 5, com três repetições, compreendendo duas cultivares (Conquista e Monsoy 8400 e cinco estádios de corte (R2, R3, R4, R5 e R6. As cultivares afetaram apenas o rendimento de proteína do feno. Os maiores rendimentos de matéria seca, feno e proteínas foram obtidos com os cortes nos estádios R4, R5 e R6. Quanto ao rendimento de proteína bruta, verificou-se queda após o corte em R5. Esta redução, provavelmente, associada a efeito de estresse hídrico e ou diluição da proteína na matéria seca. O estádio de desenvolvimento R4 apresentou a melhor opção de corte para fenação, independentemente da cultivar. A cultivar Monsoy 8400 apresentou maior rendimento de proteína bruta que a cultivar Conquista.

  4. Determining a pool - type reactor fuel policy; Recherche d'une politique de gestion du combustible d'une pile piscine

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    Refuelling the 10 to 15 MW pool type reactor considered here will occur frequently (some 10 elements every 3 to 4 weeks). It is therefore necessary to determine the most economic fuel policy. This study proposes to define a strategy that will make it possible to decide on the number and characteristics of the shipment containers, as well as on the means of storage, so as to reduce the risks as much as possible should the basic parameters of the study vary. Among these parameters, the respective influence of which is investigated, chemical reprocessing costs play a vital part. Two examples of optimum fuel management are given according to whether the reprocessing charges applied are those of the old or of the 1961 U.S. AEC base charges for reprocessing highly enriched irradiated fuel. (authors) [French] Les renouvellements du combustible de la pile piscine de 10 a 15 MW consideree ici, seront frequents (quelques 10 elements toutes les 3 ou 4 semaines). Il est donc important de connaitre les meilleures conditions de gestion economique de ce combustible. Cette etude se propose de definir une strategie permettant de choisir le nombre et les caracteristiques des containers de transport, ainsi que les moyens de stockage, de facon a minimiser les risques si les parametres de base de l'etude varient. Parmi ces parametres dont l'influence respective est successivement examinee, le cout du traitement chimique joue un role fondamental. Deux exemples d'optimum de gestion sont donnes selon que le tarif de retraitement est conforme a l'ancien tarif de l'U.S. AEC ou au tarif 1961 de traitement des combustibles irradies tres enrichis. (auteurs)

  5. En torno a una estética del deporte

    OpenAIRE

    Javier Olivera Betrán*

    2006-01-01

    Aunque el término estética tiene varios significados especiales, en el contexto de las ciencias sociales se emplea preferentemente para designar el conjunto de investigaciones que están relacionadas con las artes. Estética es el estudio del comportamiento y la experiencia del hombre al crear arte, al percibir y comprender el arte y al ser influido por el arte. El arte es el medio que promueve una fusión del individuo con el todo, el hombre para rebasar los límites que le impone la individuali...

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Arndt, D.; Schmidt, W.

    1989-01-01

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

  8. Spatial kinetics in nuclear reactor systems. Chapter 4

    International Nuclear Information System (INIS)

    Owens, D.H.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  10. A estética como factor promotor da aprendizagem

    Directory of Open Access Journals (Sweden)

    Fátima Maria BEZERRA BARBOSA

    2012-02-01

    Full Text Available La reflexión filosófica sobre el arte, lo hermoso, lo bueno, lo verdadero y su capacidad de promover el desarrollo humano es un tema que remonta a Platón en la cultura occidental. Considerando que la educación debe ser entendida como un fenómeno integral, que abarca no sólo las dimensiones intelectuales sino también las dimensiones sensibles, la intuición y el raciocinio, las capacidades imaginativas y creativas, el «saber ser» y el «saber estar» parejas del «conocer» y del «saber hacer», pensamos que es incuestionable la importancia de la estética en el fenómeno educativo. El objetivo de este trabajo es demostrar la importancia de la educación estética como motor fundamental de desarrollo humano y, en este sentido, como catalizador de la motivación y del aprendizaje, para que, a través de la imaginación y de la creatividad, sea posible la innovación en la construcción de vías alternativas para un mundo más humano y placentero.

  11. A new high performance research reactor

    International Nuclear Information System (INIS)

    Abbate, Pablo M.

    2002-01-01

    A contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000 between Australia authorities and INVAP from Argentina. Since then the detailed design has been completed, an application for a construction license was made in May 2001 and the construction authorisation was issued on 4 th April 2002. This paper explains the safety philosophy embedded into the design together with the approach taken for main elements of the design and their relation to the proposed applications of the reactor. Also information is provided on the suit of neutron beam facilities and irradiation facilities being constructed. Finally it is presented an outline of the project management organisation, project planing and schedule. (author)

  12. Rotating reactors : a review

    NARCIS (Netherlands)

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

    2013-01-01

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

  13. Flow analysis in a supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Oh, C.H.; Kochan, R.J.; Beller, J.M.

    1996-01-01

    Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-12-15

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Setiyanto; Sembiring, Tagor M.; Pinem, Surian

    2007-01-01

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

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

    OpenAIRE

    Zambrano Flores, Johanna Vanessa

    2013-01-01

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

  18. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1988-01-01

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

  19. Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

    International Nuclear Information System (INIS)

    Morrone, A.; Brussalis, W.G.

    1975-01-01

    The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-12-31

    Radiodosimetria de la Autoridad Regulatoria Nuclear (ARN) utiliza la columna rapida del reactor para el disenio, calibracion y puesta a punto de dosimetros de criticidad asi como de los metodos de evaluacion rapida de las dosis absorbida por el personal en el caso de accidentes. Con este objeto se irradiaron hojuelas de Au(1), Au bajo Cd e In(2), para estimar los flujos neutronicos absolutos termico, epitermico y rapido respectivamente en la posicion de irradiacion. Esta estimacion es tanto mas precisa cuanto mejor se conocen las respuestas de los distintos materiales de los detectores al espectro de neutrones presente. Esto a su vez requiere el conocimiento previo de dicho espectro (dependencia detallada del flujo neutronico con la energia) en la posicion que se analiza. En este trabajo se presenta el calculo neutronico que fue oportunamente requerido al Grupo de Reactores de Investigacion y Conjuntos Criticos de la ARN, con el objeto de determinar el espectro neutronico en la posicion de irradiacion rapida del RA-1. Se ha evaluado el espectro neutronico en diferentes posiciones del reactor, utilizando dos lineas de calculo diferentes y considerando, a los fines de este analisis, una potencia de 40 kW. Se ha representado al reactor y al recinto que lo aloja con un modelo simplificado, unidimensional, como un conjunto de regiones circulares concentricas. Se muestran los graficos de los flujos rapido y termico (con corte en 0,4 eV) en funcion de la distancia al centro del reactor. Se muestra asimismo el grafico del flujo neutronico (en n/cm{sup 2}.seg.eV) en funcion de la energia en la posicion de irradiacion rapida. Se consignan tambien los valores del flujo (en n/cm{sup 2}.seg.eV) en funcion de la energia en otras posiciones tipicas, asi como los valores equivalentes de los flujos integrados (en n/cm{sup 2}.seg). ((1) De acuerdo con la reaccion Au{sup 197}(n, {gamma})Au{sup 198}, con una seccion eficaz de {sigma}{sub 0}=98.8b para neutrones termicos. (2) De acuerdo con la

  1. Cermet fuel reactors

    International Nuclear Information System (INIS)

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

    1987-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-15

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

  3. Enhanced degradation of p-chlorophenol in a novel pulsed high voltage discharge reactor.

    Science.gov (United States)

    Bian, Wenjuan; Ying, Xiangli; Shi, Junwen

    2009-03-15

    The yields of active specie such as ozone, hydrogen peroxide and hydroxyl radical were all enhanced in a novel discharge reactor. In the reactor, the original formation rate of hydroxyl radical was 2.27 x 10(-7) mol L(-1)s(-1), which was about three times than that in the contrast reactor. Ozone was formed in gas-phase and was transferred into the liquid. The characteristic of mass transfer was better in the novel reactor than that in the contrast reactor, which caused much higher ozone concentration in liquid. The dissociation of hydrogen peroxide was more evident in the former, which promoted the formations of hydroxyl radical. The p-chlorophenol (4-CP) degradation was also enhanced. Most of the ozone transferred into the liquid and hydrogen peroxide generated by discharge could be utilized by the degradation process of 4-CP. About 97% 4-CP was removed in 36 min discharge in the novel reactor. Organic acids such as formic, acetic, oxalic, propanoic and maleic acid were generated and free chloride ions were released in the degradation process. With the formation of organic acid, the pH was decreased and the conductivity was increased.

  4. Enhanced degradation of p-chlorophenol in a novel pulsed high voltage discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bian Wenjuan [School of Chemistry and Chemical Engineering, Soochow University, Suzhou 215123 (China)], E-mail: bwenjuan@suda.edu.cn; Ying Xiangli; Shi Junwen [School of Chemistry and Chemical Engineering, Soochow University, Suzhou 215123 (China)

    2009-03-15

    The yields of active specie such as ozone, hydrogen peroxide and hydroxyl radical were all enhanced in a novel discharge reactor. In the reactor, the original formation rate of hydroxyl radical was 2.27 x 10{sup -7} mol L{sup -1} s{sup -1}, which was about three times than that in the contrast reactor. Ozone was formed in gas-phase and was transferred into the liquid. The characteristic of mass transfer was better in the novel reactor than that in the contrast reactor, which caused much higher ozone concentration in liquid. The dissociation of hydrogen peroxide was more evident in the former, which promoted the formations of hydroxyl radical. The p-chlorophenol (4-CP) degradation was also enhanced. Most of the ozone transferred into the liquid and hydrogen peroxide generated by discharge could be utilized by the degradation process of 4-CP. About 97% 4-CP was removed in 36 min discharge in the novel reactor. Organic acids such as formic, acetic, oxalic, propanoic and maleic acid were generated and free chloride ions were released in the degradation process. With the formation of organic acid, the pH was decreased and the conductivity was increased.

  5. Enhanced degradation of p-chlorophenol in a novel pulsed high voltage discharge reactor

    International Nuclear Information System (INIS)

    Bian Wenjuan; Ying Xiangli; Shi Junwen

    2009-01-01

    The yields of active specie such as ozone, hydrogen peroxide and hydroxyl radical were all enhanced in a novel discharge reactor. In the reactor, the original formation rate of hydroxyl radical was 2.27 x 10 -7 mol L -1 s -1 , which was about three times than that in the contrast reactor. Ozone was formed in gas-phase and was transferred into the liquid. The characteristic of mass transfer was better in the novel reactor than that in the contrast reactor, which caused much higher ozone concentration in liquid. The dissociation of hydrogen peroxide was more evident in the former, which promoted the formations of hydroxyl radical. The p-chlorophenol (4-CP) degradation was also enhanced. Most of the ozone transferred into the liquid and hydrogen peroxide generated by discharge could be utilized by the degradation process of 4-CP. About 97% 4-CP was removed in 36 min discharge in the novel reactor. Organic acids such as formic, acetic, oxalic, propanoic and maleic acid were generated and free chloride ions were released in the degradation process. With the formation of organic acid, the pH was decreased and the conductivity was increased

  6. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

    1996-01-01

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

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

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

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

  8. Farewell to a reactor

    International Nuclear Information System (INIS)

    Skanborg, P.

    1976-01-01

    Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

  9. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    International Nuclear Information System (INIS)

    Essadki, A.H.; Gourich, B.; Vial, Ch.; Delmas, H.; Bennajah, M.

    2009-01-01

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm 2 , but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H 2 microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  10. Behavior of the future LHC magnet protection diodes irradiated in a nuclear reactor at 4.6 K with intermediate annealing

    International Nuclear Information System (INIS)

    Berland, V.; Hagedorn, D.; Gerstenberg, H.

    1996-01-01

    In the framework of the LHC project at CERN, the effects of radiation on the electrical characteristics of epitaxial diodes for superconducting magnet protection were studied. The diodes were exposed to an irradiation dose up to 50 kGy and a neutron fluence of 10 15 n/cm 2 with intermediate thermal annealing each 10 kGy dose steps in the Technical University of Munich reactor at 4.6 K

  11. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  12. IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

    International Nuclear Information System (INIS)

    Cunningham, Mitchel E.; Turnbull, J.A.

    2003-01-01

    Description - Objectives - Results: The U.S. Nuclear Regulatory Commission (NRC) conducted a series of thermal-hydraulic and cladding mechanical deformation tests in the National Research Universal (NRU) reactor at the Chalk River National Laboratory in Canada. The objective of these tests was to perform simulated loss-of-coolant-accident (LOCA) experiments using full-length light-water reactor fuel rods to study mechanical deformation, flow blockage, and coolability. Three phases of a LOCA (i.e., heat-up, reflood, and quench) were performed in situ using nuclear fissioning to simulate the low-level decay power during a LOCA after shutdown. All tests used PWR-type, non-irradiated fuel rods. Provided here is information on two materials tests, MT-6A and MT-4, which PNNL considers the better characterized for the purposes of setting up computer cases. The NRU reactor is a heterogeneous, thermal, tank-type research reactor. It has a power level of 135 MWth and is heavy-water moderated and cooled. The coolant has an inlet temperature of 310 K at a pressure of 0.65 MPa. The MT tests were conducted in a specially designed test train to supply the specified coolant conditions of flowing steam, stagnant steam, and then reflood. Typical instrumentation for the MT tests included fuel centerline thermocouples, cladding inner surface thermocouples, cladding outer surface thermocouples, rod internal gas pressure transducers or pressure switches, coolant channel steam probes, and self-powered neutron detectors. This instrumentation allowed for determining rupture times and cladding temperature. The test rods for the LOCA cases in the NRU reactor were irradiated in flowing steam prior to the transient, stagnant steam during the transient and prior to reflood, and then reflood conditions to complete the transient. Both cladding inner surface and outer surface temperatures were measured, in addition to coolant temperatures. However, only cladding inner surface temperatures were

  13. Sobre a estética platônica

    Directory of Open Access Journals (Sweden)

    Luis Felipe Bellintani Ribeiro

    2007-04-01

    Full Text Available O texto começa pela consideração da aparente posição desfavorável da estética, entendida tanto como o domínio da arte, quanto como o relativo à sensação, perante a metafísica platônica. A consideração posterior, porém, do caráter ontológico do critério metafísico, que, em tese, poria a arte e a sensação a seu reboque, ensejou, mais que uma reabilitação, uma inversão: a metafísica platônica é que seria fundada numa estética.

  14. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

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

  15. 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)

  16. 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

  17. Performance of a UASB reactor treating coffee wet wastewater

    International Nuclear Information System (INIS)

    Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

    2014-01-01

    The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-11-15

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

  19. Efficient Synthesis of Ethanol from CH4 and Syngas on a Cu-Co/TiO2 Catalyst Using a Stepwise Reactor

    Science.gov (United States)

    Zuo, Zhi-Jun; Peng, Fen; Huang, Wei

    2016-10-01

    Ethanol synthesis from CH4 and syngas on a Cu-Co/TiO2 catalyst is studied using experiments, density functional theory (DFT) and microkinetic modelling. The experimental results indicate that the active sites of ethanol synthesis from CH4 and syngas are Cu and CoO, over which the ethanol selectivity is approximately 98.30% in a continuous stepwise reactor. DFT and microkinetic modelling results show that *CH3 is the most abundant species and can be formed from *CH4 dehydrogenation or through the process of *CO hydrogenation. Next, the insertion of *CO into *CH3 forms *CH3CO. Finally, ethanol is formed through *CH3CO and *CH3COH hydrogenation. According to our results, small particles of metallic Cu and CoO as well as a strongly synergistic effect between metallic Cu and CoO are beneficial for ethanol synthesis from CH4 and syngas on a Cu-Co/TiO2 catalyst.

  20. Neutron source for a reactor

    International Nuclear Information System (INIS)

    Kobayashi, Hiromasa.

    1975-01-01

    Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-08-01

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

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

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

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

  3. Production of 48V in a nuclear reactor via secondary tritons

    International Nuclear Information System (INIS)

    Siri, S.; Cohen, I.M.

    2009-01-01

    The production of 48 V in a nuclear reactor, induced on titanium by tritons generated from the 6 Li(n, t) 4 He reaction, and eventually 7 Li(n, n't) 4 He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from 48 V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

  4. Investigation of slightly forced buoyant flow in a training reactor

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.; Por, G.

    2001-01-01

    A measurement based on the temperature noise analysis method was carried out in the Training Reactor of the Budapest University of Technology and Economics. The main goals were the estimation of the flow velocity immediately above the reactor core and investigation of the thermal-hydraulical conditions of the reactor, mainly in the core. Subsequently 2D and 3D computations were carried out with the aid of the code CFX- 4.3. The main objective of the 2D calculation was to clarify the thermal-hydraulical conditions of the whole reactor tank with a reasonable computing demand. It was also necessary to accomplish 3D numerical investigations of the reactor core and the space above since three dimensional effects of the flow could only be studied in this way. In addition, obtaining certain boundary conditions of the 3D computations was another significant aim of the 2D investigations. It is important that the results of the noise analysis and the operational measuring system of the reactor gave us a basis for verifying our computations.(author)

  5. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

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

  6. Molten salt reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Simon, N.; Renault, C.

    2014-01-01

    Molten salt reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. The principle of this reactor is very innovative: the nuclear fuel is dissolved in the coolant which allows the online reprocessing of the fuel and the online recovery of the fission products. A small prototype: the Molten Salt Reactor Experiment (MSRE - 8 MWt) was operating a few years in the sixties in the USA. The passage towards a fast reactor by the suppression of the graphite moderator leads to the concept of Molten Salt Fast Reactor (MSFR) which is presently studied through different European projects such as MOST, ALISIA and EVOL. Worldwide the main topics of research are: the adequate materials resisting to the high level of corrosiveness of the molten salts, fuel salt reprocessing, the 3-side coupling between neutron transport, thermohydraulics and thermo-chemistry, the management of the changing chemical composition of the salt, the enrichment of lithium with Li 7 in the case of the use of lithium fluoride salt and the use of MSFR using U 233 fuel (thorium cycle). The last part of the article presents a preliminary safety analysis of the MSFR. (A.C.)

  7. The integral fast reactor concept

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  9. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

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

  10. Status of French reactors

    International Nuclear Information System (INIS)

    Ballagny, A.

    1997-01-01

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

  11. CO2 photoreduction using NiO/InTaO4 in optical-fiber reactor for renewable energy

    NARCIS (Netherlands)

    Wang, Zhen-Yi; Chou, Hung-Chi; Wu, Jeffrey C.S.; Tsai, Din Ping; Mul, Guido

    2010-01-01

    The photocatalytic reduction of CO2 into fuels provides a direct route to produce renewable energy from sunlight. NiO loaded InTaO4 photocatalyst was prepared by a sol–gel method. Aqueous-phase CO2 photoreduction was performed in a quartz reactor to search for the highest photoactivity in a series

  12. The Jules Horowitz Reactor (JHR), a European Material Testing Reactor (MTR), with extended experimental capabilities

    International Nuclear Information System (INIS)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

    2003-01-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10 14 ncm -2 s -1 and a fast flux of 6,4.10 14 ncm -2 s -1 , it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

  13. What occurred in the reactors

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko

    2013-01-01

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

  14. 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

  15. 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

  16. 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

  17. A model for reliability avaliation of the electrical supply source of the 1A3 and 1A4 control rods assemblies of Angra I reactor

    International Nuclear Information System (INIS)

    Yang, T.

    1978-01-01

    The reliability of the electrical power supply to the 4.16KV buses for the safety system operation of a nuclear power plant was studied. Particularly, Angra Unit I system was focused. Initially, reliability of each electrical supply source was estimated. Using a probabilistic approach based on the Markov processes, the system reliability was evaluated in terms of frequency and duration of loss of power supply and of the system failure probability evolution when one or more sources remained unavailable. Based on these results, certain reactor operating rules were proposed concerning later shutdown of the plant without compromising the nuclear reactor safety. A sensitivity analysis was also performed to show the different reliability parameter influences on final results. This analysis showed that the diesel system performs an important role in the power supply for a nuclear power plant [pt

  18. 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

  19. Plutonium-239 production rate study using a typical fusion reactor

    International Nuclear Information System (INIS)

    Faghihi, F.; Havasi, H.; Amin-Mozafari, M.

    2008-01-01

    The purpose of the present paper is to compute fissile 239 Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m 2 ) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate 239 Pu production rate. Produced 239 Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, 239 Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type

  20. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    International Nuclear Information System (INIS)

    Guigon, B.; Vacelet, H.; Dornbusch, D.

    2000-01-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs from activation analysis to power reactor fuel qualification. In this paper the main characteristics of the Jules Horowitz Reactor are presented. Safety criteria are explained. Finally, merits and disadvantages of UMo compared to the standard U 3 Si 2 fuel are discussed. (author)

  1. Kinetic analysis of sub-prompt-critical reactor assemblies

    International Nuclear Information System (INIS)

    Das, S.

    1992-01-01

    Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

  2. Reactor building for a nuclear reactor

    International Nuclear Information System (INIS)

    Haidlen, F.

    1976-01-01

    The invention concerns the improvement of the design of a liner, supported by a latticed steel girder structure and destined for guaranteeing a gastight closure for the plant compartments in the reactor building of a pressurized water reactor. It is intended to provide the steel girder structure on their top side with grates, being suited for walking upon, and to hang on their lower side diaphragms in modular construction as a liner. At the edges they may be sealed with bellows in order to avoid thermal stresses. The steel girder structure may at the same time serve as supports for parts of the steam pipe. (RW) [de

  3. Neurexin-1β Binding to Neuroligin-1 Triggers the Preferential Recruitment of PSD-95 versus Gephyrin through Tyrosine Phosphorylation of Neuroligin-1

    Directory of Open Access Journals (Sweden)

    Grégory Giannone

    2013-06-01

    Full Text Available Adhesion between neurexin-1β (Nrx1β and neuroligin-1 (Nlg1 induces early recruitment of the postsynaptic density protein 95 (PSD-95 scaffold; however, the associated signaling mechanisms are unknown. To dissociate the effects of ligand binding and receptor multimerization, we compared conditions in which Nlg1 in neurons was bound to Nrx1β or nonactivating HA antibodies. Time-lapse imaging, fluorescence recovery after photobleaching, and single-particle tracking demonstrated that in addition to aggregating Nlg1, Nrx1β binding stimulates the interaction between Nlg1 and PSD-95. Phosphotyrosine immunoblots and pull-down of gephyrin by Nlg1 peptides in vitro showed that Nlg1 can be phosphorylated at a unique tyrosine (Y782, preventing gephyrin binding. Expression of Nlg1 point mutants in neurons indicated that Y782 phosphorylation controls the preferential binding of Nlg1 to PSD-95 versus gephyrin, and accordingly the formation of inhibitory and excitatory synapses. We propose that ligand-induced changes in the Nlg1 phosphotyrosine level control the balance between excitatory and inhibitory scaffold assembly during synapse formation and stabilization.

  4. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  5. FLICA-4 (version 1). A computer code for three dimensional thermal analysis of nuclear reactor cores

    International Nuclear Information System (INIS)

    Raymond, P.; Allaire, G.; Boudsocq, G.; Caruge, D.; Gramont, T. de; Toumi, I.

    1995-01-01

    FLICA-4 is a thermal-hydraulic computer code, developed at the French Atomic Energy Commission (CEA) for three-dimensional steady-state or transient two-phase flow, and aimed at design and safety thermal analysis of nuclear reactor cores. It is available for various UNIX workstations and CRAY computers under UNICOS.It is based on four balance equations which include three balance equations for the mixture and a mass balance equation for the less concentrated phase which allows for the calculation of non equilibrium flows such as sub-cooled boiling and superheated steam. A drift velocity model takes into account the velocity unbalance between phases. The equations are solved using a finite volume numerical scheme. Typical running time, specific features (coupling with other codes) and auxiliary programs are presented. 1 tab., 9 refs

  6. 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)

  7. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  8. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

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

  9. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

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

  10. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  11. Industrial Ultrasonic Inspection of Stainless-Steel Claddings for the EL4 Reactor; Controle Industriel par Ultrasons des Gaines en Acier Inoxydable du Reacteur EL4; Promyshlennyj kontrol' obolochechnykh trub iz nerzhaveyushchej stali reaktora dlya EL4 s pomoshch'yu ul'trazvukovogo metoda; Metodos Ultrasonicos para Control Industrial de las Vainas de Acero Inoxidable del Reactor EL4

    Energy Technology Data Exchange (ETDEWEB)

    Prot, A. C.; Foulquoer, H. E.; Peyrot, J. P. [Centre d' Etudes Nucleaires de Saclay (France)

    1965-09-15

    Improved reactor performance requires the use of accurately fabricated and carefully inspected components. One inspection relates to the quality of the cladding tubes, whose mechanical reliability is essential for economic reactor operation. The choice and development of a method is a difficult matter and the authors explain the main factors involved. Once the choice has been made and the method has been developed in the laboratory, two new problems arise: Adaptation to meet industrial requirements; and The need to reconcile the quality standards attainable with the manufacturing process at any given stage and the somewhat arbitrarily defined specifications for the finished product. In practice, this involves a statistical study of batches of tubes from various sources and their classification in relation to more or less strict thresholds. The number of tubes which have to be inspected is much larger than originally expected. This has led to the design of an automatic inspection device geared both to the output rates involved and to the requirements of the type of inspection adopted; the latter are generally mechanical and impose particularly careful product fabrication. These various characteristics are now embodied in a device whose capacity can already easily meet the requirements of a fuel-element production line. The potentialities of the device are closely dependent on the characteristics of the inspection equipment used, especially the performances of the electronic part of ultrasonic inspection instruments and of the transducers. This study shows that standard equipment is not very suitable and that immediate thought should be given to special instruments for this type of inspection. (author) [French] L'accroissement des performances des reacteurs necessite l'utilisation de materiaux finement elabores et soigneusement controles. L'un des aspects de ce controle est celui de la qualite des tubes de gainage utilises, dont la tenue mecanique est un facteur

  12. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  13. Seismic analysis of a large pool-type LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Wang, C.Y.; Gvildys, J.

    1989-01-01

    This paper describes the seismic study of a 450-MWe liquid metal reactor (LMR) under 0.3-g SSE ground excitation. Two calculations were performed using the new design configuration. They deal with the seismic response of the reactor vessel, the guard vessel and support skirt, respectively. In both calculations, the stress and displacement fields at important locations of those components are investigated. Assessments are also made on the elastic and inelastic structural capabilities for other beyond-design basis seismic loads. Results of the reactor vessel analysis reveal that the maximum equivalent stress is only about half of the material yield stress. For the guard vessel and support skirt, the stress level is very small. Regarding the analysis if inelastic structural capability, solutions of the Newmark-Hall ductility modification method show that the reactor vessel can withstand seismics with ground ZPAs ranging from 1.015 to 1.31 g, which corresponds to 3.37 to 4.37 times the basic 0.3-g SSE. Thus, the reactor vessel and guard vessel are strong enough to resist seismic loads. 4 refs., 10 figs., 5 tabs

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

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

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

  15. Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

    International Nuclear Information System (INIS)

    Ogawa, Yuichi

    1998-01-01

    Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

  16. EST! best practices

    Science.gov (United States)

    2000-10-05

    This booklet presents the eighteen winners in a competition for projects that promote environmentally sustainable transportation (EST). These range from car sharing to hydrogen fueled buses to public transit promotion campaign. Each project is presen...

  17. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

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

  18. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  19. Some aspects related to the management of maintenance for a TRIGA research reactor In Romania

    International Nuclear Information System (INIS)

    Vieru, G.

    2003-01-01

    Safety management for a nuclear research reactor involves 'good dependability management' of operations activities, such as: reliability, availability, maintainability and maintenance support. In order to evaluate the safety management aspects intended to be applied by research reactor management, the performance dependability indicators and their impact on reactor availability and reactor safety have to be established. The document ISO 9000-4/IEC 300-1 'Dependability Management' (1995), describes five internationally agreed indicators of reactor equipment dependability. Each of them can be used for corrective maintenance or for preventive maintenance, such as: I 1 - equipment Maintenance Frequency; I 2 - equipment Maintenance Effort; I 3 - equipment Maintenance Downtime Factor; I 4 - equipment Maintenance Contribution to the System Function Downtime Factor; I 5 - equipment Maintenance Contribution to the reactor Capability Loss Factor. This paper presents an evaluation of those 5 mentioned indicators with reference to the primary circuit of the INR's TRIGA research reactor and conclusion. The analyzed period was between 1994- 1999. It is to be noted that this type of analysis is performed for the first time for a research reactor. (author)

  20. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

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

  1. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

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

  2. Biochemical and Structural Analysis of Hormone-sensitive Lipase Homolog EstE7: Insight into the Stabilized Dimerization of HSL-Homolog Proteins

    International Nuclear Information System (INIS)

    Nam, Ki Hyun; Park, Sung Ha; Lee, Won Ho; Hwang, Kwang Yeon

    2010-01-01

    Hormone sensitive lipase (HSL) plays a major role in energy homeostasis and lipid metabolism. Several crystal structures of HSL-homolog proteins have been identified, which has led to a better understanding of its molecular function. HSLhomolog proteins exit as both monomer and dimer, but the biochemical and structural basis for such oligomeric states has not been successfully elucidated. Therefore, we determined the crystal structure of HSL-homolog protein EstE7 from a metagenome library at 2.2 A resolution and characterized the oligomeric states of EstE7 both structurally and biochemically. EstE7 protein prefers the dimeric state in solution, which is supported by its higher enzymatic activity in the dimeric state. In the crystal form, EstE7 protein shows two-types of dimeric interface. Specifically, dimerization via the external β8-strand occurred through tight association between two pseudosymmetric folds via salt bridges, hydrogen bonds and van der Waals interactions. This dimer formation was similar to that of other HSL-homolog protein structures such as AFEST, BEFA, and EstE1. We anticipate that our results will provide insight into the oligomeric state of HSLhomolog proteins

  3. Turismo y estética formativa

    Directory of Open Access Journals (Sweden)

    Clara Inés Sánchez Arciniegas

    2002-11-01

    Full Text Available ¿Existe alguna armonía entre el conocimiento y la estética? Sin duda alguna el hombre con conocimiento y sin virtud de la estética es un autómata; este necesita de las diferentes manifestaciones en las que se encuentra plasmada y necesita recordarlas continuamente. La estética como doctrina de conocimiento no es reciente, en el siglo xviii cuando por primera vez se define teóricamente hasta nuestros días en la Universidad actual. Esta última aparentemente menos evidente debido a las nuevas formas de aprendizaje con herramientas tecnológicas muy avanzadas y contraria a la imagen del universitario que hace dos siglos como complemento a su educación se distraía a través del Gran Tour, partida que tiene el turismo para fomentar el goce estético. ¿Cuál es entonces la relación entre distracción y formación? Se debe entonces reflexionar y comprender que el conocimiento interdisciplinario y la metodología debe procurar en los hombres actitudes adecuadas frente a la cultura y el arte a través del conocimiento.

  4. Comparison of simulated and measured quantities of a duplex reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koskela, M.; Kajava, M. [ABB Marine, Helsinki (Finland)

    1997-12-31

    The purpose of this article is to illustrate the use of an analog simulator as a design tool when designing new power electric equipment. The purpose of simulation is to predict the functionality of electrical equipment to be constructed. Duplex reactor is an electromagnetic device designed to reduce voltage harmonics and short circuit currents in the ship electrical network. In this report a comparison between simulated and measured electrical quantities of a duplex reactor has been made. The purpose of the measurements was to show the correct functioning of the reactor. The simulation results and the measured waveforms corresponds well to each other. (orig.) 4 refs.

  5. Aqueous homogeneous suspension reactor project. Report over the 4th quarter and the year 1974

    Energy Technology Data Exchange (ETDEWEB)

    1975-07-01

    The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described.

  6. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  7. A estética política das mídias locativas

    Directory of Open Access Journals (Sweden)

    Lucia Santaella

    2008-04-01

    Full Text Available Las prácticas de acceso propiciadas por los dispositivos móviles conectados al Sistema de Posicionamiento Global (GPS están haciendo surgir un nuevo espacio social de combinaciones inextricables entre lo virtual (el ciberespacio y los ambientes físicos en los que se mueve nuestro cuerpo biológico. Este trabajo busca presentar las nuevas modalidades de comunicación, conexión e interacción que están apareciendo bajo el nombre de “medios de comunicación locativos”. Éstos son procesos que están dando origen a una estética políticamente orientada, cuyas principales características se discutirán en el artículo.

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

    Directory of Open Access Journals (Sweden)

    S. N. Rokhman

    2015-04-01

    been carried out for the molten salt fuel LiF-BeF2-ThF4-UF4 on a Small Mobile-Molten Salt Reactor (SM-SMR. The core configurations and operating temperature should be adjusted in using the new fuel in order to get the calculated keff and CR (conversion ratio are > 1 in the fraction of 0.5% 233U, 20% 232Th, 28% Li, 51.5% Be. After obtained that keff and CR close to 1, then the analysis of changes in the Th to Be and Be to Li are carried out, it indicates the changes of keff and CR. Then the 233U fraction is varied between 0.5–0.46% to obtain the condition keff > 1 and CR > 1. To determine the temperature coefficient of reactivity (αT,the temperature of core is changed about +25K dan +50K. To determine the void reactivity coefficient (αV, fuel density is reduced to 90%. The result shows that the reduction of Th causes the decrease of CR and increase of keff due to the number fertile material is less. The addition of Be to Li will make the keff is increase and the CR is decrease, because the macroscopic absorption cross section of Li is greater than Be. From the five 233U composition in the ranges 0.5–0.46%, the calculated keff and CR varies in the range of 1.00001 – 1.00327 and 1.00016 – 1.00731, respectively. For power peaking factor (PPF, the calculation results give the value in the range of 2.4311 - 2.4714. However, for the safety parameters, the negative temperature reactivity coefficient (αT and negative void reactivity CR (αV in the range of 4.972×10-5 – 5.909×10-5 and 2.596×10-2 - 2,8287×10-2 ∆k/k/K, respectively. It can be concluded that the SM-MSR core has negative value for those reactivity for all fractions, so the core fulfill the safety criteria and inherent safety. Keywords: small mobile molten salt reactor (SM-MSR, LiF-BeF2-ThF4-UF4 fuel, inherent safety, temperature coefficient reactivity, void coefficient reactivity.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  10. Design analysis and microprocessor based control of a nuclear reactor

    International Nuclear Information System (INIS)

    Sabbakh, N.J.

    1988-01-01

    The object of this thesis is to design and test a microprocessor based controller, to a simulated nuclear reactor system. The mathematical model that describes the dynamics of a typical nuclear reactor of one group of delayed neutrons approximations with temperature feedback was chosen. A digital computer program has been developed for the design and analysis of a simulated model based on the concept of state-variable feedback in order to meet a desired system response with maximum overshoot of 3.4% and setting time of 4 sec. The state variable feedback coefficients are designed for the continuous system, then an approximation is used to obtain in the state variable feedback vector for the discrete system. System control was implemented utilizing Direct Digital Control (DDC) of a nuclear reactor simulated model through a control algorithm that was performed by means of a microprocessor based system. The controller performance was satisfactorily tested by exciting the reactor system with a transient reactivity disturbance and by a step change in power demand. Direct digital control, when implemented on a microprocessor adds versatility, flexibility in system design with the added advantage of possible use of optimal control algorithms. 6 tabs.; 30 figs.; 46 refs.; 6 apps

  11. 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

  12. Public information circular for shipments of irradiated reactor fuel. Revision 4

    International Nuclear Information System (INIS)

    1984-06-01

    This publication is the fifth in a series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. This publication contains basically three kinds of information: (1) routes recently approved (18 months) by the Commission for the shipment of irradiated reactor fuel; (2) information regarding any safeguards-significant incidents that may be (to date none have) reported during shipments along such routes; and (3) cumulative amounts of material shipped

  13. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

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

  14. The European Pressurized Water Reactor. A safe and competitive solution for future energy needs

    International Nuclear Information System (INIS)

    Leverenz, R.; Gerhard, L.; Goebel, A.

    2004-01-01

    The European Pressurized Water Reactor - the EPR - is a PWR in the 1600 MW class. Its design is based on experience feedback from several thousand reactors x years of light water reactor operation worldwide, primarily those incorporating the most recent technologies: the French N4 and the German KONVOI reactors. It is an evolutionary design that ensures continuity in the mastery of PWR technology, minimizing the risk for the customer. (author)

  15. Pyrolysis of Rubber in a Screw Reactor

    Science.gov (United States)

    Lozhechnik, A. V.; Savchin, V. V.

    2016-11-01

    On the basis of an analysis of thermal methods described in the literature and from the results of experimental investigations of steam conversion, the authors have developed and created a facility for thermal processing of rubber waste. Rubber crumb was used as the raw material; the temperature in the reactor was 500°C; nitrogen, steam, and a mixture of light hydrocarbons (noncondensable part of pyrolysis products) represented the working medium. The pyrolysis yielded 36-38% of a solid fraction, 54-56% of a liquid hydrocarbon fraction, and 6-9% of noncondensable gases. Changes in the composition of the gas mixture have been determined at different stages of processing. Gas chromatography of pyrolysis gases has shown that the basic gases produced by pyrolysis are H2 and hydrocarbons C2H4, C3H6, C3H8, C4H8, C2H6, C3H6O2, and C4H10, and a small amount of H2S, CO, and CO2. Noncondensable gases will be used as a fuel to heat the reactor and to implement the process.

  16. Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

    1991-01-01

    A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

  17. Plutonium-239 production rate study using a typical fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Havasi, H.; Amin-Mozafari, M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

    2008-05-15

    The purpose of the present paper is to compute fissile {sup 239}Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m{sup 2}) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate {sup 239}Pu production rate. Produced {sup 239}Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, {sup 239}Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type.

  18. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

    1997-01-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  19. 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.)

  20. Degradation pathway of malachite green in a novel dual-tank photoelectrochemical catalytic reactor

    International Nuclear Information System (INIS)

    Diao, Zenghui; Li, Mingyu; Zeng, Fanyin; Song, Lin; Qiu, Rongliang

    2013-01-01

    Highlights: • A novel dual-tank photoelectrochemical catalytic reactor was designed. • Malachite green degraded in bipolar double-effect mode. • Salt bridge replaced by a cation exchange membrane in the reactor. • Degradation pathways of malachite green in the cathode and anode tanks were similar. -- Abstract: A novel dual-tank photoelectrochemical catalytic reactor was designed to investigate the degradation pathway of malachite green. A thermally formed TiO 2 /Ti thin film electrode was used as photoanode, graphite was used as cathode, and a saturated calomel electrode was employed as the reference electrode in the reactor. In the reactor, the anode and cathode tanks were connected by a cation exchange membrane. Results showed that the decolorization ratio of malachite green in the anode and cathode was 98.5 and 96.5% after 120 min, respectively. Malachite green in the two anode and cathode tanks was oxidized, achieving the bipolar double effect. Malachite green in both the anode and cathode tanks exhibited similar catalytic degradation pathways. The double bond of the malachite green molecule was attacked by strong oxidative hydroxyl radicals, after which the organic compound was degraded by the two pathways into 4,4-bis(dimethylamino) benzophenone, 4-(dimethylamino) benzophenone, 4-(dimethylamino) phenol, and other intermediate products. Eventually, malachite green was degraded into oxalic acid as a small molecular organic acid, which was degraded by processes such as demethylation, deamination, nitration, substitution, addition, and other reactions

  1. Synthesis of Struvite using a Vertical Canted Reactor with Continuous Laminar Flow Process

    Science.gov (United States)

    Sutiyono, S.; Edahwati, L.; Muryanto, S.; Jamari, J.; Bayuseno, A. P.

    2018-01-01

    Struvite is a white crystalline that is chemically known as magnesium ammonium phosphorus hexahydrate (MgNH4PO4·6H2O). It can easily dissolve in acidic conditions and slightly soluble in neutral and alkaline conditions. In industry, struvite forms as a scale deposit on a pipe with hot flow fluid. However, struvite can be used as fertilizer because of its phosphate content. A vertical canted reactor is a promising technology for recovering phosphate levels in wastewater through struvite crystallization. The study was carried out with the vertical canted reactor by mixing an equimolar stock solution of MgCl2, NH4OH, and H3PO4 in 1: 1: 1 ratio. The crystallization process worked with the flow rate of three stock solution entering the reactor in the range of 16-38 ml/min, the temperature in the reactor is worked on 20°, 30°, and 40°C, while the incoming air rate is kept constant at 0.25 liters/min. Moreover, pH was maintained at a constant value of 9. The struvite crystallization process run until the steady state was reached. Then, the result of crystal precipitates was filtered and dried at standard temperature room for 48 hours. After that, struvite crystals were stored for the subsequent analysis by Scanning Electron Microscope (SEM) and XRD (X-Ray Diffraction) method. The use of canted reactor provided the high pure struvite with a prismatic crystal morphology.

  2. Medical Radioisotopes Production Without A Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Keur, H.

    2010-05-15

    This report is answering the key question: Is it possible to ban the use of research reactors for the production of medical radioisotopes? Chapter 2 offers a summarized overview on the history of nuclear medicine. Chapter 3 gives an overview of the basic principles and understandings of nuclear medicine. The production of radioisotopes and its use in radiopharmaceuticals as a tracer for imaging particular parts of the inside of the human body (diagnosis) or as an agent in radiotherapy. Chapter 4 lists the use of popular medical radioisotopes used in nuclear imaging techniques and radiotherapy. Chapter 5 analyses reactor-based radioisotopes that can be produced by particle accelerators on commercial scale, other alternatives and the advantages of the cyclotron. Chapter 6 gives an overview of recent developments and prospects in worldwide radioisotopes production. Chapter 7 presents discussion, conclusions and recommendations, and is answering the abovementioned key question of this report: Is it possible to ban the use of a nuclear reactor for the production of radiopharmaceuticals? Is a safe and secure production of radioisotopes possible?.

  3. Biological biogas upgrading capacity of a hydrogenotrophic community in a trickle-bed reactor

    International Nuclear Information System (INIS)

    Rachbauer, Lydia; Voitl, Gregor; Bochmann, Günther; Fuchs, Werner

    2016-01-01

    Highlights: • Data on long term operation of a system supplied with real biogas are presented. • Ex-situ biological methanation is feasible for biogas upgrading. • Gas quality obtained complies with strictest direct grid injection criteria. • Biomethane can act as flexible storage for renewable surplus electricity. - Abstract: The current study reports on biological biogas upgrading by means of hydrogen addition to obtain biomethane. A mesophilic (37 °C) 0.058 m"3 trickle-bed reactor with an immobilized hydrogenotrophic enrichment culture was operated for a period of 8 months using a substrate mix of molecular hydrogen (H_2) and biogas (36–42% CO_2). Complete CO_2 conversion (> 96%) was achieved up to a H_2 loading rate of 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d, corresponding to 2.3 h gas retention time. The optimum H_2/CO_2 ratio was determined to be between 3.67 and 4.15. CH_4 concentrations above 96% were achieved with less than 0.1% residual H_2. This gas quality complies even with tightest standards for grid injection without the need for additional CO_2 removal. If less rigid standards must be fulfilled H_2 loading rates can be almost doubled (10.95 versus 6.5 m_n"3 H_2/m"3_r_e_a_c_t_o_r _v_o_l_. × d) making the process even more attractive. At this H_2 loading the achieved methane productivity was 2.52 m_n"3 CH_4/m"3_r_e_a_c_t_o_r _v_o_l_. × d. In terms of biogas this corresponds to an upgrading capacity of 6.9 m_n"3 biogas/m"3_r_e_a_c_t_o_r _v_o_l_. × d. The conducted experiments demonstrate that biological methanation in an external reactor is well feasible for biogas upgrading under the prerequisite that an adequate H_2 source is available.

  4. Safety equipment in a reactor

    International Nuclear Information System (INIS)

    Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

    1976-01-01

    Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

  5. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    1995-02-01

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

  6. Fusion reactor development: A review

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

  7. Investigation on the radiation damage behavior of various alloys in a fusion reactor using thorium molten salt

    International Nuclear Information System (INIS)

    Ubeyli, Mustafa; Demir, Teyfik

    2008-01-01

    In fusion reactors, one of the most important problems is the need for the frequent change of the first wall material during the reactor's operation due to the radiation damage induced by high energetic particles, especially fusion neutrons coming from fusion plasma. In order to solve this problem, in HYLIFE-II fusion reactor design, a liquid wall between the fusion plasma and first wall is used. This study presents the radiation damage behaviors of candidate structural materials (9Cr-2WVTa, V-4Cr-4Ti and W-5Re alloys) considered to be used in fusion reactors to determine the optimum thickness of the liquid wall in HYLIFE-II fusion reactor. In the liquid wall, a thorium molten salt consisting of 75%LiF-23%ThF 4 -2% 233 UF 4 was used. Calculations were carried out with respect to the variable liquid wall thickness and for an operation period of 30 years. Numerical results related to atomic displacement and helium generation damage pointed out that the liquid wall thickness should be at least 42, 66 and 81 cm for the materials, W-5Re, 9Cr-2WVTa, V-4Cr-4Ti, respectively in order not to exceed relevant damage limits after a reactor operation of 30 years

  8. Reactor FaceMap Tool: A modern graphics tool for displaying reactor data

    International Nuclear Information System (INIS)

    Roberts, J.C.

    1991-01-01

    A prominent graphical user interface in reactor physics applications at the Savannah River Site is the reactor facemap display. This is a two dimensional view of a cross section of a reactor. In the past each application which needed a facemap implemented its own version. Thus, none of the code was reused, the facemap implementation was hardware dependent and the user interface was different for each facemap. The Reactor FaceMap Tool was built to solve these problems. Through the use of modern computing technologies such as X Windows, object-oriented programming and client/server technology the Reactor FaceMap Tool has the flexibility to work in many diverse applications and the portability to run on numerous types of hardware

  9. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  10. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given. (author)

  11. Research reactor instrumentation and control technology. Report of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-10-01

    The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

  12. Design of first reactor protection system prototype for C A R E M reactor

    International Nuclear Information System (INIS)

    Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

    2006-01-01

    In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

  13. Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

  14. The count variance-covariance matrix in a critical reactor

    International Nuclear Information System (INIS)

    Carloni, F.; Giovannini, R.

    1984-01-01

    The present paper deals with a critical reactor containing a set of neutron detectors operating one at time in different time intervals. The analysis makes use of the Kolmogorov backward formalism for the branching processes, in the framework of the one-velocity, point reactor model, explicitly taking into account the six groups of delayed neutrons. The expression of the mean value, the covariance of the counting distribution are reported. The list of the Fortran 4. subroutine CRITIC which computes these moments is also reported

  15. International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    1973-01-01

    The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

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

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

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

  17. PAVE: Program for assembling and viewing ESTs

    Directory of Open Access Journals (Sweden)

    Bomhoff Matthew

    2009-08-01

    Full Text Available Abstract Background New sequencing technologies are rapidly emerging. Many laboratories are simultaneously working with the traditional Sanger ESTs and experimenting with ESTs generated by the 454 Life Science sequencers. Though Sanger ESTs have been used to generate contigs for many years, no program takes full advantage of the 5' and 3' mate-pair information, hence, many tentative transcripts are assembled into two separate contigs. The new 454 technology has the benefit of high-throughput expression profiling, but introduces time and space problems for assembling large contigs. Results The PAVE (Program for Assembling and Viewing ESTs assembler takes advantage of the 5' and 3' mate-pair information by requiring that the mate-pairs be assembled into the same contig and joined by n's if the two sub-contigs do not overlap. It handles the depth of 454 data sets by "burying" similar ESTs during assembly, which retains the expression level information while circumventing time and space problems. PAVE uses MegaBLAST for the clustering step and CAP3 for assembly, however it assembles incrementally to enforce the mate-pair constraint, bury ESTs, and reduce incorrect joins and splits. The PAVE data management system uses a MySQL database to store multiple libraries of ESTs along with their metadata; the management system allows multiple assemblies with variations on libraries and parameters. Analysis routines provide standard annotation for the contigs including a measure of differentially expressed genes across the libraries. A Java viewer program is provided for display and analysis of the results. Our results clearly show the benefit of using the PAVE assembler to explicitly use mate-pair information and bury ESTs for large contigs. Conclusion The PAVE assembler provides a software package for assembling Sanger and/or 454 ESTs. The assembly software, data management software, Java viewer and user's guide are freely available.

  18. PAVE: program for assembling and viewing ESTs.

    Science.gov (United States)

    Soderlund, Carol; Johnson, Eric; Bomhoff, Matthew; Descour, Anne

    2009-08-26

    New sequencing technologies are rapidly emerging. Many laboratories are simultaneously working with the traditional Sanger ESTs and experimenting with ESTs generated by the 454 Life Science sequencers. Though Sanger ESTs have been used to generate contigs for many years, no program takes full advantage of the 5' and 3' mate-pair information, hence, many tentative transcripts are assembled into two separate contigs. The new 454 technology has the benefit of high-throughput expression profiling, but introduces time and space problems for assembling large contigs. The PAVE (Program for Assembling and Viewing ESTs) assembler takes advantage of the 5' and 3' mate-pair information by requiring that the mate-pairs be assembled into the same contig and joined by n's if the two sub-contigs do not overlap. It handles the depth of 454 data sets by "burying" similar ESTs during assembly, which retains the expression level information while circumventing time and space problems. PAVE uses MegaBLAST for the clustering step and CAP3 for assembly, however it assembles incrementally to enforce the mate-pair constraint, bury ESTs, and reduce incorrect joins and splits. The PAVE data management system uses a MySQL database to store multiple libraries of ESTs along with their metadata; the management system allows multiple assemblies with variations on libraries and parameters. Analysis routines provide standard annotation for the contigs including a measure of differentially expressed genes across the libraries. A Java viewer program is provided for display and analysis of the results. Our results clearly show the benefit of using the PAVE assembler to explicitly use mate-pair information and bury ESTs for large contigs. The PAVE assembler provides a software package for assembling Sanger and/or 454 ESTs. The assembly software, data management software, Java viewer and user's guide are freely available.

  19. EST Vocabulary Instruction

    Directory of Open Access Journals (Sweden)

    Célia D.S. Bell

    2012-05-01

    Full Text Available This study aims at contributing to the investigation on the instruction of EST (English for Science and Technology vocabulary, in terms of receptive use of the language. It evaluates the effectiveness of two teaching approaches to the acquisition of vocabulary. The first approach consisted of teaching vocabulary through the use of dictionaries, where the words were merely translated into the learners’ L1 or defined in the target language thus promoting superficial level of word processing. The second approach employed activities promoting deep level of word processing. Data were analysed quantitatively. Results indicated that the two approaches seem to have some equipotentiality, as far as EST vocabulary is concerned.

  20. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  2. Proceedings of the 4th international symposium on material testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Suzuki, Masahide

    2012-03-01

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

  3. Proceedings of the 4th international symposium on material testing reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-03-15

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

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

    International Nuclear Information System (INIS)

    Ardani

    2011-01-01

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

  5. Improvement of research reactor sustainability

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

    2010-01-01

    The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)

  6. Exploiting EST databases for the development and characterization of EST-SSR markers in castor bean (Ricinus communis L.

    Directory of Open Access Journals (Sweden)

    Yang Jun-Bo

    2010-12-01

    Full Text Available Abstract Background The castor bean (Ricinus communis L., a monotypic species in the spurge family (Euphorbiaceae, 2n = 20, is an important non-edible oilseed crop widely cultivated in tropical, sub-tropical and temperate countries for its high economic value. Because of the high level of ricinoleic acid (over 85% in its seed oil, the castor bean seed derivatives are often used in aviation oil, lubricants, nylon, dyes, inks, soaps, adhesive and biodiesel. Due to lack of efficient molecular markers, little is known about the population genetic diversity and the genetic relationships among castor bean germplasm. Efficient and robust molecular markers are increasingly needed for breeding and improving varieties in castor bean. The advent of modern genomics has produced large amounts of publicly available DNA sequence data. In particular, expressed sequence tags (ESTs provide valuable resources to develop gene-associated SSR markers. Results In total, 18,928 publicly available non-redundant castor bean EST sequences, representing approximately 17.03 Mb, were evaluated and 7732 SSR sites in 5,122 ESTs were identified by data mining. Castor bean exhibited considerably high frequency of EST-SSRs. We developed and characterized 118 polymorphic EST-SSR markers from 379 primer pairs flanking repeats by screening 24 castor bean samples collected from different countries. A total of 350 alleles were identified from 118 polymorphic SSR loci, ranging from 2-6 per locus (A with an average of 2.97. The EST-SSR markers developed displayed moderate gene diversity (He with an average of 0.41. Genetic relationships among 24 germplasms were investigated using the genotypes of 350 alleles, showing geographic pattern of genotypes across genetic diversity centers of castor bean. Conclusion Castor bean EST sequences exhibited considerably high frequency of SSR sites, and were rich resources for developing EST-SSR markers. These EST-SSR markers would be particularly

  7. Reinforced confinement in a nuclear reactor

    International Nuclear Information System (INIS)

    Norman, H.

    1988-01-01

    The present invention concerns a nuclear reactor containing a reactor core, a swimming pool space that is filled and pressurized with a neutron-absorbing solution, a reactor tank, at least one heat exchanger, at least one inlet line, at least one return line and at least one circulation pump, where the said reactor tank is confined in the said swimming pool space and designed to be cooled with the aid of relatively pure water, which is fed by means of the said at least one circulating pump to the said reactor tank from the said heat exchanger via the said at least one inlet line and is returned to the heat exchanger via the said at least one return line. The problem that is to be solved by the invention is to design a reactor of the above type in such a way that a complete confinement of the primary circuit of the reactor is achieved at relatively low extra cost. This problem is solved by providing the reactor with a special confinement space that confines the heat exchanger, but not the reactor tank, with the confinement space and the swimming pool space being fashioned in the same concrete body

  8. Presentación de algunos conceptos básicos del cálculo infinitesimal a partir de unos principios no estándar

    Directory of Open Access Journals (Sweden)

    de la Torre Cuesta, C.

    1999-01-01

    Full Text Available En el artículo “A NEW APPROACH TO NONSTANDARD ANALYSIS” del profesor Manuel Suárez Fernández, se dan unos principios no estándar compatibles con los axiomas de la teoría de conjuntos ZFC . Estos principios son: PRINCIPIO 1 : Existe algún número natural no estándar. PRINCIPIO 2: Si n, m son números naturales, n estándar y m no estándar entonces nA es un conjunto estándar y no vacío entonces existe algún elemento estándar de A. PRINCIPIO 4: Si r es un real limitado entonces existe un número real estándar que denotamos or y llamamos parte estándar de r, tal que or-r es un infinitésimo. En esta comunicación tras dar la definición no estándar de sucesión regular y de sucesión convergente para sucesiones estándar se demuestran algunas propiedades a partir de los mencionados principios.

  9. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

  10. Study of the formation and of the distribution of dissolved gases and hydrogen peroxide in water from a swimming-pool reactor (triton) (1961); Etude de la formation et de la repartition des gaz dissous et de l'eau oxygenee dans l'eau d'un reacteur piscine (triton) (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Chenouard, J; Rozenberg, J; Dolle, L; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    In order to determine experimentally the amount of radiolysis in the swimming-pool reactor Triton, direct measurements have been made of the quantity of radiolysis gas and hydrogen peroxide in the water, at the entry and exit of the core. The concentration distribution of these gases in the reactor was also determined. An explanation is given as to why no gases evolution is seen in the swimming-pool reactors of the C.E.A. The overall amount of radiolysis is zero, and a simple interpretation of this result is possible. The real amount of radiolysis occurring in the reactor core can be calculated. This is in satisfactory agreement with certain measurement mad elsewhere. (authors) [French] Pour determiner experimentalement le taux de radiolyse dans la pile piscine Triton, des mesures directes de la quantite de gaz de radiolyse et d'eau oxygenee dans l'eau a l'entree et a la sortie du coeur ont ete faites. La repartition de la concentration de ces gaz dans la piscine a egalement ete determinee. On explique pourquoi aucun degagement gazeux n'est observe dans les piles piscines du CE.A. Le taux de radiolyse global est nul, et une interpretation simple de ce resultat est possible. Un taux de radiolyse reel dans le coeur du reacteur peut etre calcule. Celui-ci est en accord satisfaisant avec certaines determinations faites ailleurs. (auteurs)

  11. Prospect of realizing nuclear fusion reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

  12. Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

    International Nuclear Information System (INIS)

    Feng, K.M.; Huang, J.H.

    1994-01-01

    The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

  13. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  14. Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

    International Nuclear Information System (INIS)

    Chang, Si Young; Ha, Chung Woo

    1988-01-01

    An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

  16. Reactor physics aspects of burning actinides in a nuclear reactor

    International Nuclear Information System (INIS)

    Hage, W.; Schmidt, E.

    1978-01-01

    A short review of the different recycling strategies of actinides other than fuel treated in the literature, is given along with nuclear data requirements for actinide build-up and transmutation studies. The effects of recycling actinides in a nuclear reactor on the flux distribution, the infinite neutron multiplication factor, the reactivity control system, the reactivity coefficients and the delayed neutron fraction are discussed considering a notional LWR or LMFBR as an Actinide Trasmutaton Reactor. Some operational problems of Actinide Transmutation reactors are mentioned, which are caused by the α-decay heat and the neutron sources of Actinide Target Elements

  17. Emergency core cooling system for a fast reactor

    International Nuclear Information System (INIS)

    Johnson, H.G.; Madsen, R.N.

    1976-01-01

    The main heat transport system for a liquid-metal-cooled nuclear reactor is constructed with elevated piping and guard vessels or pipes around all components of the system below the elevation of the elevated piping so the head developed by the pumps at emergency motor speed will be unsufficient to lift the liquid-metal-coolant over the top of the guard tanks or pipes or out of the elevated piping in the event of a loss-of-coolant accident. In addition, inlet downcomers to the reactor vessel are contained within guard standpipes having a clearance volume as small as practicable. 4 claims, 2 drawing figures

  18. A history of effluent releases from the Texas A and M University reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bates, E F; Neff, R D; Sandel, P S; Schoenbucher, B [Texas A and M University (United States)

    1974-07-01

    Since 1966 records of radioactive effluents releases from the Texas A and M University Research Reactor have been compiled. These data include particulate activity, noble gases, and liquid effluent releases. Particulate activity releases with half-lives greater than eight days were negligible and are not included in this presentation. Conversion from an MTR plate reactor to a TRIGA fueled reactor was completed in August 1968. Records of effluent releases of Argon-4l and liquids for the past, five years are summarized, in this presentation. These release data are compared to the current limits specified: in 10 CPR 20 and the limits appearing in proposed Appendix.

  19. A reduced fidelity model for the rotary chemical looping combustion reactor

    International Nuclear Information System (INIS)

    Iloeje, Chukwunwike O.; Zhao, Zhenlong; Ghoniem, Ahmed F.

    2017-01-01

    Highlights: • Methodology for developing a reduced fidelity rotary CLC reactor model is presented. • The reduced model determines optimal reactor configuration that meets design and operating requirements. • A 4-order of magnitude reduction in computational cost is achieved with good prediction accuracy. • Sensitivity studies demonstrate importance of accurate kinetic parameters for reactor optimization. - Abstract: The rotary chemical looping combustion reactor has great potential for efficient integration with CO_2 capture-enabled energy conversion systems. In earlier studies, we described a one-dimensional rotary reactor model, and used it to demonstrate the feasibility of continuous reactor operation. Though this detailed model provides a high resolution representation of the rotary reactor performance, it is too computationally expensive for studies that require multiple model evaluations. Specifically, it is not ideal for system-level studies where the reactor is a single component in an energy conversion system. In this study, we present a reduced fidelity model (RFM) of the rotary reactor that reduces computational cost and determines an optimal combination of variables that satisfy reactor design requirements. Simulation results for copper, nickel and iron-based oxygen carriers show a four-order of magnitude reduction in simulation time, and reasonable prediction accuracy. Deviations from the detailed reference model predictions range from 3% to 20%, depending on oxygen carrier type and operating conditions. This study also demonstrates how the reduced model can be modified to deal with both optimization and design oriented problems. A parametric study using the reduced model is then applied to analyze the sensitivity of the optimal reactor design to changes in selected operating and kinetic parameters. These studies show that temperature and activation energy have a greater impact on optimal geometry than parameters like pressure or feed fuel

  20. Development and production of an oligonucleotide MuscleChip: use for validation of ambiguous ESTs

    Directory of Open Access Journals (Sweden)

    Lanfranchi Gerolamo

    2002-10-01

    Full Text Available Abstract Background We describe the development, validation, and use of a highly redundant 120,000 oligonucleotide microarray (MuscleChip containing 4,601 probe sets representing 1,150 known genes expressed in muscle and 2,075 EST clusters from a non-normalized subtracted muscle EST sequencing project (28,074 EST sequences. This set included 369 novel EST clusters showing no match to previously characterized proteins in any database. Each probe set was designed to contain 20–32 25 mer oligonucleotides (10–16 paired perfect match and mismatch probe pairs per gene, with each probe evaluated for hybridization kinetics (Tm and similarity to other sequences. The 120,000 oligonucleotides were synthesized by photolithography and light-activated chemistry on each microarray. Results Hybridization of human muscle cRNAs to this MuscleChip (33 samples showed a correlation of 0.6 between the number of ESTs sequenced in each cluster and hybridization intensity. Out of 369 novel EST clusters not showing any similarity to previously characterized proteins, we focused on 250 EST clusters that were represented by robust probe sets on the MuscleChip fulfilling all stringent rules. 102 (41% were found to be consistently "present" by analysis of hybridization to human muscle RNA, of which 40 ESTs (39% could be genome anchored to potential transcription units in the human genome sequence. 19 ESTs of the 40 ESTs were furthermore computer-predicted as exons by one or more than three gene identification algorithms. Conclusion Our analysis found 40 transcriptionally validated, genome-anchored novel EST clusters to be expressed in human muscle. As most of these ESTs were low copy clusters (duplex and triplex in the original 28,000 EST project, the identification of these as significantly expressed is a robust validation of the transcript units that permits subsequent focus on the novel proteins encoded by these genes.

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

    OpenAIRE

    Cuevas-Rodríguez, Germán; Tejero Monzón, Iñaki

    2003-01-01

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

  2. Development of an innovative reflector drive mechanism using magnetic repulsion force for 4S reactor

    International Nuclear Information System (INIS)

    Tsuji, K.; Watanabe, M.; Inagaki, H.; Nishikawa, A.; Takahashi, H.; Wakamatsu, M.; Matsumiya, H.; Nishiguchi, Y.

    2001-01-01

    A small sized fast reactor 4S: (Super Safe Small and Simple) which has a core of 10 - 30 years life time is controlled by reflectors. The reflector is required to be risen at very low speed to make up for the reactivity swing during operation. This report shows the development of an innovative reflector drive mechanism using magnetic repulsion force that can move at a several micrometer per one step. This drive mechanism has a passive shut down capability, and can eliminate reflector drive line. (author)

  3. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    Tashanii, Ahmed Ali

    2010-01-01

    Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

  4. The experimental nuclear reactor: AQUILON; Le reacteur nucleaire experimental: AQUILON

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Koechlin, J C; Moreau, J M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

  5. BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of program or function: The system of codes can be used to solve nuclear reactor core static neutronics and reactor history exposure problems. BOLD/VENTURE-4: First order perturbation and time-dependent sensitivity theories can be applied. Control rod positioning may be modeled explicitly and refueling treated with repositioning and recycle. Special capability is coded to model the continuously fueled core and to solve the importance and dominant harmonics problems. The modules of the code system are: VENTNEUT: VENTURE neutronics module; DRIVER and CONTRL: Control module; BURNER: Exposure calculation for reactor core analysis; FILEDTOR: File editor; INPROSER: Input processor; EXPOSURE: BURNER code module; REACRATE: Reaction rate calculation; CNTRODPO: Control rod positioning; FUELMANG: Fuel management positioning and accounting; PERTUBAT: Perturbation reactivity importance analyses; sensitivity analysis; DEPTHMOD: Static and time-dependent perturbation sensitivity analysis. The special processors are: DVENTR: Handles the input to the VENTURE module; DCMACR: Converts CITATION macroscopic cross sections to microscopic cross sections; DCRSPR: Produces input for the CROSPROS module; DUTLIN: Adds or replaces problem input data without exiting the program; DENMAN: Repositions fuel; DMISLY: Miscellaneous tasks. Standard interface files between modules are binary sequential files that follow a standardized format. VENTURE-PC: The microcomputer version is a subset of the mainframe version. The modules and special processors which are not part of VENTURE-PC are: REACRATE, CNTRODPO, PERTUBAT, FUELMANG, DEPTHMOD, DMISLY. 2 - method of solution: BOLD-VENTURE-4: The neutronics problems are solved by applying the multigroup diffusion theory representation of neutron transport applying an over-relaxation inner iteration, outer iteration scheme. Special modeling is used or source correction is done during iteration to solve importance and harmonics problems. No

  6. Reactor safety in Eastern Europe. Proceedings

    International Nuclear Information System (INIS)

    1995-02-01

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

  7. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-09-01

    A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

  8. Fusion reactor materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

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

  9. Fast reactor irradiation effects on fracture toughness of Si_3N_4 in comparison with MgAl_2O_4 and yttria stabilized ZrO_2

    International Nuclear Information System (INIS)

    Tada, K.; Watanabe, M.; Tachi, Y.; Kurishita, H.; Nagata, S.; Shikama, T.

    2016-01-01

    Fracture toughness of silicon nitride (Si_3N_4), magnesia-alumina spinel (MgAl_2O_4) and yttria stabilized zirconia (8 mol%Y_2O_3–ZrO_2) was evaluated by the Vickers-indentation technique after the fast reactor irradiation up to 55 dpa (displacement per atom) at about 700 °C in the Joyo. The change of the fracture toughness by the irradiation was correlated with nanostructural evolution by the irradiation, which was examined by transmission electron microscopy. The observed degradation of fracture toughness in Si_3N_4 is thought to be due to the relatively high density of small-sized of the irradiation induced defects, which should be resulted from a large amount of transmutation gases of hydrogen and helium. Observed improvement of fracture toughness in MgAl_2O_4 was due to the blocking of crack propagation by the antiphase boundaries. The radiation effects affected the fracture toughness of yttria stabilized zirconia at 55 dpa, suggesting that the generated high density voids would affect the propagation of cracks. - Highlights: • Si_3N_4, MgAl_2O_4 and YSZ were neutron irradiated up to 55dpa around 700 °C in the Joyo. • They are candidate ceramics for the inert matrices of nuclear fuels in the fast reactors. • The irradiation enhanced the fracture toughness of MgAl_2O_4 and YSZ, while degraded that of Si_3N_4. • The toughness changes were correlated with radiation induced defects and transmutation gases.

  10. Evaluación estética de seis tipos de coronas para dientes primarios

    Directory of Open Access Journals (Sweden)

    Héctor Alejandro Ramírez Peña

    2017-03-01

    Full Text Available Objetivo: Evaluar las preferencias estéticas en relación con el color y la forma de coronas primarias utilizadas para dientes incisivos superiores primarios, mediante la realización de una encuesta a miembros de la Academia Mexicana de Odontología Pediatrica (AMOP. Material y Métodos: Se establecieron seis grupos de estudio con seis coronas diferentes: grupo 1, coronas de zirconia EZ-Pedo; grupo 2, coronas de zirconia NuSmile Zr; grupo 3, coronas estéticas hechas en el consultorio; grupo 4, coronas de fundas de celuloide; grupo 5, coronas estéticas prefabricadas NuSmile signature; y grupo 6, coronas estéticas fenestradas. Se llevaron a cabo encuestas con la finalidad de conocer las preferencias estéticas de estas diferentes coronas, con la finalidad de conocer cuál es la mejor opción para su uso en el consultorio dental. Resultados: Noventa miembros de la AMOP realizaron una encuesta válida, y se determinó que el grupo 4 fue el mejor evaluado, seguido de los grupos 2, 5, 1, 6 y 3. Se identificaron diferencias significativas entre los diferentes grupos. Conclusiones: Las coronas de fundas de celuloide fueron seleccionadas como mejor alternativa de uso en los dientes primarios anteriores, por parte de los miembros de la AMOP; asimismo, se consideró a las coronas de zirconia como una buena opción terapéutica. Es recomendable que se implemente el tratamiento estético en dientes primarios, para realizar un tratamiento integral.

  11. 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.)

  12. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  13. Biodegradation of 4-bromophenol by Arthrobacter chlorophenolicus A6 in batch shake flasks and in a continuously operated packed bed reactor.

    Science.gov (United States)

    Sahoo, Naresh Kumar; Pakshirajan, Kannan; Ghosh, Pranab Kumar

    2014-04-01

    The present study investigated growth and biodegradation of 4-bromophenol (4-BP) by Arthrobacter chlorophenolicus A6 in batch shake flasks as well as in a continuously operated packed bed reactor (PBR). Batch growth kinetics of A. chlorophenolicus A6 in presence of 4-BP followed substrate inhibition kinetics with the estimated biokinetic parameters value of μ max = 0.246 h(-1), K i = 111 mg L(-1), K s  = 30.77 mg L(-1) and K = 100 mg L(-1). In addition, variations in the observed and theoretical biomass yield coefficient and maintenance energy of the culture were investigated at different initial 4-BP concentration. Results indicates that the toxicity tolerance and the biomass yield of A. chlorophenolicus A6 towards 4-BP was found to be poor as the organism utilized the substrate mainly for its metabolic maintenance energy. Further, 4-BP biodegradation performance by the microorganism was evaluated in a continuously operated PBR by varying the influent concentration and hydraulic retention time in the ranges 400-1,200 mg L(-1) and 24-7.5 h, respectively. Complete removal of 4-BP was achieved in the PBR up to a loading rate of 2,276 mg L(-1) day(-1).

  14. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

  15. Production of {sup 48}V in a nuclear reactor via secondary tritons

    Energy Technology Data Exchange (ETDEWEB)

    Siri, S. [Comision Nacional de Energia Atomica, Centro Atomico Ezeiza, Gerencia de Capacitacion, Quimica Nuclear y Ciencias de la Salud, Ezeiza, Buenos Aires (Argentina); Cohen, I.M. [Univ. Tecnologica Nacional, Dept. de Ingenieria Quimica, Buenos Aires (Argentina)

    2009-07-01

    The production of {sup 48}V in a nuclear reactor, induced on titanium by tritons generated from the {sup 6}Li(n, t){sup 4} He reaction, and eventually {sup 7}Li(n, n't){sup 4}He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from {sup 48}V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

  16. A gamma heating calculation methodology for research reactor application

    International Nuclear Information System (INIS)

    Lee, Y.K.; David, J.C.; Carcreff, H.

    2001-01-01

    Gamma heating is an important issue in research reactor operation and fuel safety. Heat deposition in irradiation targets and temperature distribution in irradiation facility should be determined so as to obtain the optimal irradiation conditions. This paper presents a recently developed gamma heating calculation methodology and its application on the research reactors. Based on the TRIPOLI-4 Monte Carlo code under the continuous-energy option, this new calculation methodology was validated against calorimetric measurements realized within a large ex-core irradiation facility of the 70 MWth OSIRIS materials testing reactor (MTR). The contributions from prompt fission neutrons, prompt fission γ-rays, capture γ-rays and inelastic γ-rays to heat deposition were evaluated by a coupled (n, γ) transport calculation. The fission product decay γ-rays were also considered but the activation γ-rays were neglected in this study. (author)

  17. El leviatán estético

    Directory of Open Access Journals (Sweden)

    Jorge Peñuela

    2011-05-01

    Full Text Available En este ensayo exploro algunos argumentos de Hannah Arendt, con el fin de mostrar las razones que explican el odio que se manifestaba en las discusiones artísticas de comienzos del siglo XX, odio que persiste aún en muchos escenarios artísticos actuales. Con Arendt sostengo que este fenómeno no está superado, porque el odio es la herramienta que despliega el Capital, para amalgamar a todos los desheredados que ha creado. Contextualizo en la práctica artística colombiana el siguiente argumento de Arendt: el capitalismo creó una amalgama de frustrados sociales, políticos y estéticos que destruye la esfera pública como lugar para el acontecer de la política, amalgama que denominó populacho. Introduzco la categoría de populacho estético, para hablar del odio que ha animado las prácticas artísticas en Colombia desde 1940.

  18. EST Express: PHP/MySQL based automated annotation of ESTs from expression libraries.

    Science.gov (United States)

    Smith, Robin P; Buchser, William J; Lemmon, Marcus B; Pardinas, Jose R; Bixby, John L; Lemmon, Vance P

    2008-04-10

    Several biological techniques result in the acquisition of functional sets of cDNAs that must be sequenced and analyzed. The emergence of redundant databases such as UniGene and centralized annotation engines such as Entrez Gene has allowed the development of software that can analyze a great number of sequences in a matter of seconds. We have developed "EST Express", a suite of analytical tools that identify and annotate ESTs originating from specific mRNA populations. The software consists of a user-friendly GUI powered by PHP and MySQL that allows for online collaboration between researchers and continuity with UniGene, Entrez Gene and RefSeq. Two key features of the software include a novel, simplified Entrez Gene parser and tools to manage cDNA library sequencing projects. We have tested the software on a large data set (2,016 samples) produced by subtractive hybridization. EST Express is an open-source, cross-platform web server application that imports sequences from cDNA libraries, such as those generated through subtractive hybridization or yeast two-hybrid screens. It then provides several layers of annotation based on Entrez Gene and RefSeq to allow the user to highlight useful genes and manage cDNA library projects.

  19. EST Express: PHP/MySQL based automated annotation of ESTs from expression libraries

    Directory of Open Access Journals (Sweden)

    Pardinas Jose R

    2008-04-01

    Full Text Available Abstract Background Several biological techniques result in the acquisition of functional sets of cDNAs that must be sequenced and analyzed. The emergence of redundant databases such as UniGene and centralized annotation engines such as Entrez Gene has allowed the development of software that can analyze a great number of sequences in a matter of seconds. Results We have developed "EST Express", a suite of analytical tools that identify and annotate ESTs originating from specific mRNA populations. The software consists of a user-friendly GUI powered by PHP and MySQL that allows for online collaboration between researchers and continuity with UniGene, Entrez Gene and RefSeq. Two key features of the software include a novel, simplified Entrez Gene parser and tools to manage cDNA library sequencing projects. We have tested the software on a large data set (2,016 samples produced by subtractive hybridization. Conclusion EST Express is an open-source, cross-platform web server application that imports sequences from cDNA libraries, such as those generated through subtractive hybridization or yeast two-hybrid screens. It then provides several layers of annotation based on Entrez Gene and RefSeq to allow the user to highlight useful genes and manage cDNA library projects.

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

    OpenAIRE

    Mendoza Sépulveda, César Camilo

    2013-01-01

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

  1. 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

  2. International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

  3. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    International Nuclear Information System (INIS)

    Kuscu, Ozlem Selcuk; Sponza, Delia Teresa

    2009-01-01

    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m 3 day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m 3 day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m 3 day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m 3 day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m 3 day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m 3 day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from the influent, anaerobic and to

  4. activismo y estéticas en experimento

    Directory of Open Access Journals (Sweden)

    Omar Rincón

    2005-01-01

    Full Text Available Podemos comunicarlo todo: tenemos las tecnologías de la comunicación cada vez más cerca; pero no podemos contar nada, porque no tenemos nada qué decir, ni sabemos cómo decirlo. Hay que inventarlo todo, intervenir de nuevo la sociedad desde la comunicación. Es necesario convertir los medios en experiencia estética, social y política. ¿Cómo? Usando a las tecnologías para el activismo y el experimento; recurriendo a los llamado retro, porque no es lo más actual, lo más mercado, como experiencia de inspiración y emoción. Hay que volver a las estéticas, a los modos de narrar y a las historias que están inscritas en nuestras identidades locales; en nuestras sensibilidades femeninas, juveniles o sexuales; en nuestras tradiciones étnicas, o en nuestros modos urbanos. Sólo basta una minicámara para inventar el universo. En este ensayo se realiza un alegato contra los universos estético-narrativopolíticos construidos por la máquina televisiva y la homogeneidad del mercado. En este texto se argumenta a favor del activismo comunicativo, de la exploración tecnológica y de la inspiración étnica como estrategias para generar una mayor conciencia sobre el valor social de la comunicación, la intervención estético-narrativa de la sociedad y la importancia política de los medios de la gente.

  5. Group cross-sections for fast reactors; Sections efficaces de groupes pour les reacteurs a neutrons rapides; Gruppovye secheniya reaktorov na bystrykh nejtronakh; Secciones eficaces de grupos para reactores rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Zweifel, P P [University of Michigan, Ann Arbor, MI (United States); Ball, G L [Atomic Power Development Associates, Inc., Detroit, MI (United States)

    1962-03-15

    , comme c'est souvent le cas, la section efficace de groupe en terme d'integrales de resonance efficace, mais qu'il faut modifier cette definition suivant le type de schema multigroupe utilise. (author) [Spanish] La memoria discute en terminos generales las ecuaciones de difusion de grupos multiples y la forma correcta de las secciones eficaces correspondientes . En particular, demuestra que la seccion eficaz media de transporte puede expresarse con bastante precision en terminos de un promedio de recorridos libres medios. Esta magnitud es dificil de calcular porque no se puede expresar en funcion de promedios elementales ; sin embargo, se demuestran varias desigualdades que simplifican el procedimiento de determinacion de promedios. La memoria discute otros tres aspectos de las secciones eficaces de grupos que con frecuencia se ignoran, pero que pueden ser importantes al estudiar detalladamente un diseno. a) El empleo de los mismos valores medios correspondientes a las secciones eficaces de grupos para todos los reactores rapidos no se justifica si los espectros de los diferentes reactores no son similares y si las secciones eficaces varian rapidamente dentro del grupo, como ocurre a menudo. Los autores describen un metodo de iteracion, que permite obtener valores medios correctos y determinar en que medida los efectos espectrales ejercen influencia sobre los calculos de reactores. b) En los calculos de transporte (metodo S{sub n} por ejemplo), los promedios deben evaluarse en funcion del angulo y de la energia. Como el flujo no es separable en una parte angulo y en una parte energetica, es necesario proceder con sumo cuidado para evitar errores. La ecuacion S{sub n} se estudia sobre la base de un modelo sencillo, y de este estudio se deduce un criterio que puede ser de utilidad al determinar la importancia de la no-separabilida d angular en los calculos de reactores. c) Basandose en los argumentos de conservacion neutronica, se deriva una relacion de compatibilidad

  6. Corrosion of reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-01-15

    Much operational experience and many experimental results have accumulated in recent years regarding corrosion of reactor materials, particularly since the 1958 Geneva Conference on the Peaceful Uses of Atomic Energy, where these problems were also discussed. It was, felt that a survey and critical appraisal of the results obtained during this period had become necessary and, in response to this need, IAEA organized a Conference on the Corrosion of Reactor Materials at Salzburg, Austria (4-9 June 1962). It covered many of the theoretical, experimental and engineering problems relating to the corrosion phenomena which occur in nuclear reactors as well as in the adjacent circuits

  7. 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.)

  8. Estudio de seguridad del acoplamiento de un reactor de alta temperatura (VHTR) con una planta de producción de hidrógeno, análisis de los accidentes relevantes asociados

    OpenAIRE

    Moyart, Quentin

    2008-01-01

    Un estudio preliminar de seguridad dedicado al acoplamiento de un reactor nuclear de muy alta temperatura (Very High Temperature Reactor) con una planta de producción de hidrógeno (HYdrogen Production Plant) según el proceso de electrólisis a alta temperatura (High Temperature Electrolysis) se presenta en esta memoria. La primera parte de este documento está dedicada a las hipótesis de funcionamiento elegidas para el estudio, así como a la breve descripción de las dos instalaciones en el esta...

  9. Los estándares de proyectos en la implementación de los estándares internacionales de información financiera

    Directory of Open Access Journals (Sweden)

    Carlos Augusto Rincón-Soto

    2014-01-01

    Full Text Available La exigencia de la implementación de los estándares internacionales en información financiera está articulada a los términos de otros estándares internacionales, aunque no es una restricción adquirir los estándares contables sin integrar los otros estándares en Colombia, estos se acoplan de manera sistémica, en un apoyo integrador. Realizar una implementación adecuada, requiere de tomar estrategias razonables, este trabajo muestra como los estándares de proyectos ofrecen los fundamentos necesarios para dirigir y gestionar el montaje de un sistema de información contable integral, que incluya las diferentes necesidades de los usuarios.

  10. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  11. PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gas Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results

  12. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  13. Contribution to the improvement of the evaluation methods of nuclear heating in reactors by using the Monte Carlo code TRIPOLI-4

    International Nuclear Information System (INIS)

    Peron, Arthur

    2014-01-01

    Technological irradiation programs carried out in experimental reactors are crucial for the support of the current nuclear fleet in terms of study and anticipation of the behavior under irradiation of fuels and structural materials. These programs make it possible to improve the safety of the current reactors and also to study materials for the new concepts of reactors. Irradiation conditions of materials in experimental reactors must be representative of those of nuclear power plants (NPPs). One of the main advantages of material testing reactors (MTRs) is to be able to carry out instrumented irradiations by adjusting experimental parameters, in particular the neutron flux and the temperature. The control of the parameter temperature of a device irradiated in an experimental reactor requires the knowledge of the nuclear heating (source term) due to the deposition of energy of the photons and the neutrons interacting in the device. A relevant evaluation of this heating is a key data for the thermal studies of design and safety of devices. The objective of this thesis is to improve the methods of the evaluation of nuclear heating in reactors. This work consists of the development of an innovating and complete coupled neutron-photon calculation scheme (allowing to obtain the contribution of neutrons, prompt gamma and decay gamma), mainly based on the 3D, continuous energy TRIPOLI-4 Monte Carlo transport code. An experimental validation of the calculation scheme has been performed, based on calorimetry measurements carried out in the OSIRIS reactor at CEA Saclay. Sensitivity studies have been undertaken to establish the impact of various parameters on nuclear heating calculations (in particular nuclear data) and to fix the final calculation scheme to be closer to the technological irradiation aspects. The thesis work leads to an operational and predictive tool for the nuclear heating estimation, meeting the experimentation needs of research reactors and can be

  14. A new fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1986-01-01

    A new nuclear reactor design based on the fluidized bed concept is proposed. A current design utilizes spherical fuel of slightly enriched Zircaloy-clad uranium dioxide fluidized by light water under pressure. The reactor is modular in system; therefore, any size reactor can be constructed from the basic standard modul. The reactor physics calculations show that reactivity increases with porosity to a maximum value and thereafter decreases. This produces inherent safety and eliminates the need for control rods and burnable poisons. The heat transfer calculations show that the maximum power extracted from the reactor core is not limited to the material temperature limits but to the maximum mass flow of coolant, which corresponds to the desired operating porosity. Design simplicity and inherent safety make it an attractive small reactor design. (Author) [pt

  15. Reactivity feedback coefficients of a low enriched uranium fuelled material test research reactor at end-of-life

    International Nuclear Information System (INIS)

    Muhammad, Farhan

    2011-01-01

    Highlights: → The isotopic concentration in the fuel changes as soon as it starts its operation. → The neutronic properties of a reactor also change with fuel burnup. → The reactivity feedbacks at end-of-life of a material test reactor fuelled with low enriched uranium fuel are calculated. → Codes used include WIMS-D4 and CITATION. - Abstract: The reactivity feedback coefficients at end-of-life of a material test reactor fuelled with low enriched uranium fuel were calculated. The reactor used for the study was the IAEA's 10 MW benchmark reactor. Simulations were carried out to calculate the different reactivity feedback coefficients including Doppler feedback coefficient, reactivity coefficient for change of water temperature and reactivity coefficient for change of water density. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It was observed that the magnitude of all the reactivity feedback coefficients increased at end of life of the reactor by almost 2-5%.

  16. Possibilities of TWR and long life reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

    2010-01-01

    Bill Gates identified the need to switch to zero-emission energy and clarified investing in Terra Power developing the TWR (Traveling Wave Reactor) in February 2010. He also visited Toshiba developing small reactor 4S (Super Safe Small and Simple). In Japan design studies of the TWR have been conducted on the CANDLE reactor without refueling and the 4S long life reactor with maintenance free. In this feature article, the state of R and D on the TWR in Japan and IAEA's activities on small reactors without online refueling were reviewed in addition to articles on impacts of Bill Gates' investment in the TWR and state of the TWR development from an interview with John Gilleland of Terra Power. (T. Tanaka)

  17. Flow-injection spectrophotometric determination of captopril in pharmaceutical formulations using a new solid-phase reactor containing AgSCN immobilized in a polyurethane resin

    Directory of Open Access Journals (Sweden)

    Fernando Campanhã Vicentini

    2012-06-01

    Full Text Available A simple flow-injection analysis procedure was developed for determining captopril in pharmaceutical formulations employing a novel solid-phase reactor containing silver thiocyanate immobilized in a castor oil derivative polyurethane resin. The method was based on silver mercaptide formation between the captopril and Ag(I in the solid-phase reactor. During such a reaction, the SCN- anion was released and reacted with Fe3+, which generated the FeSCN2+ complex that was continuously monitored at 480 nm. The analytical curve was linear in the captopril concentration range from 3.0 × 10-4 mol L-1 to 1.1 × 10-3 mol L-1 with a detection limit of 8.0 × 10-5 mol L-1. Recoveries between 97.5% and 103% and a relative standard deviation of 2% for a solution containing 6.0 × 10-4 mol L-1 captopril (n = 12 were obtained. The sample throughput was 40 h-1 and the results obtained for captopril in pharmaceutical formulations using this procedure and those obtained using a pharmacopoeia procedure were in agreement at a 95% confidence level.Um procedimento simples de análise por injeção em fluxo foi desenvolvido para a determinação de captopril em formulações farmacêuticas empregando um novo reator em fase sólida contendo tiocianato de prata imobilizado em resina poliuretana obtida a partir de óleo de mamona. O método foi baseado na formação de um mercapto composto de prata, no reator em fase sólida, obtido entre o captopril e Ag (I imobilizada. Durante a reação, íons SCN- eram liberados e reagiam com Fe3+, gerando o complexo FeSCN2+, que foi continuamente monitorado em 480 nm. A curva analítica foi linear no intervalo de concentração de captopril entre 3,0 × 10-4 a 1,1 × 10-3 mol L-1 com um limite de detecção de 8,0 × 10-5 mol L-1. Recuperações entre 97,5-103% e desvio padrão relativo de 2% para uma solução contendo 6,0 × 10-4 mol L-1 de captopril (n = 12 foram obtidos. A frequência de amostragem foi de 40 h-1 e os resultados

  18. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  19. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  20. How many reactor accidents will there be

    International Nuclear Information System (INIS)

    Islam, S.; Lindgren, K.

    1986-01-01

    A method for calculation of the probability of nuclear accidents is described. The method is based on the use of data from reactor operating experience, i.e. there have been two major accidents [Three Mile Island and Chernobyl] during 4,000 reactor-years (cumulative operating experience). The authors argue that this method is better than the present ''technical risk assessment'' method based on the likelihood of failure of a reactor component or safety system, used by designers of nuclear reactor. (U.K.)