About neutron capture therapy method development at WWR-SM reactor in institute of Nuclear Physics of Uzbekistan Academy of Sciences

International Nuclear Information System (INIS)

Abdullaeva, G.A.; Baytelesov, S.A.; Dosimbaev, A.A.; Koblik, Yu.N.; Gritsay, O.O.

2006-01-01

Full text: Neutron capture therapy (NCT) is developing method of swellings treatment, on which specialists set one's serious hopes, as at its realization the practical possibilities of the effect on any swellings open. The essence of method is simple and lies in the fact that to the swelling enter preparation containing boron or gadolinium, which one have a large capture cross-section of the thermal and slow neutrons. Then the swelling is irradiated once with the slow (epithermal) neutron beam with fluency about 10 9 neutrons /sm 2 s for a short time and single. As a result of thermal neutrons capture by the boron (or gadolinium) nuclei secondary radiation which affecting swelling cells is emitted. NCT of oncologic diseases makes the specific demands to physical parameters of neutron beams. Now research reactors are often used for NCT. However, research reactor WWR-SM (INP, Uzbekistan AS, Tashkent) doesn't provide with the epithermal neutron beams and to develop this technique the reactor, first of all, needs for obtaining the epithermal neutron beams with energy spectrum in range from 1 eV up to 10 keV and with intensity ∼ 10 9 neutron /sm 2 s. Practically it is connected with upgrade of at least one of existed reactor channels, namely with equipping with the special equipment (filters), forming from the reactor spectrum the beam of necessary energy neutrons. It requires realization of preliminary model calculations, including calculations of capture cross-sections, of filters types and their geometrical parameters on the basis of optimal selected materials. Such calculations, as a rule, are carried out on the basis of Monte-Carlo method and designed software for calculation of nuclear reactor physical and technical characteristics [1]. In this work the calculation results of devices variants and problems discussion, related with possibility of WWR-SM reactor using for NCT are presented. (author)

Production of the sealed gamma-radiation sources of with iridium-192 radionuclide at the WWR-K research reactor

International Nuclear Information System (INIS)

Petukhov, V.K.; Chernayev, V.P.; Chabeyev, N.T.; Ermakov, E.L.; Chakrov, P.V.

2005-01-01

Full text: Conversion orientation of the WWR-K research reactor activity was established after renewal of its operation in 1997. A priority in reactor works was determined in the decision of tasks of practical use of nuclear technologies in a national economy in the next directions: in an industry, public health services and agriculture. The items of prime tasks: development and introduction of radiation technologies and manufacturing of radioisotopes for industry. This task included both scientific and technical program in the list of works of the Republican goals. At the WWR-K reactor within the framework of the this task solution the works on pilot production of the sealed sources of radioactive radiations (SSRR) with Ir-192 radionuclide for an industry of Republic of Kazakhstan were made. Organizational questions related to the Kazakhstan authority body and the regulating documentation were solved the first of all. The second stage was the development of the techniques of creating of devices providing an samples irradiation in reactor, control of sources sealing, measurements of the equivalent radiation doze from sources and high-quality support of SSRR manufacture over all technological way. At the third stage was made a little quantity SSRR with Ir-192 radionuclide, such as GIID-A1 (G6), for 'TEKOPS-660' Gammaray Projectors. This work served as experimental check of the decisions correctness, and has allowed to remove those lacks, to find out which it was possible only during direct manufacturing of radioactive sources. During performance of all these works the following was carried out: development and release of the documents and specifications regulating work on SSRR manufacture at the Institute of Nuclear Physics; personnel preparation and certification; preparation and equipment providing of reactor hot chambers by additional devices for work with irradiated iridium samples; development and manufacturing of the devices for iridium samples irradiation in

Some aspects related to radioprotection during decommissioning of the WWR-S research reactor

International Nuclear Information System (INIS)

Pantazi, Doina; Stan, Camelia

2007-01-01

Radiological safety management ensures protection of personnel, public and environment. During decommissioning of a WWR-S type research reactor, besides other specific industrial problems, radiation and/or contamination sources will be produced and their effects have to be kept under control. In any decommissioning operation that implies working in a radioactive environment, the main concern being the minimization of the total dose received by the workers. To minimize the possible dose that an individual could receive, prior entering the working area, a definite set of stages of a radiation protection plan, developed according to ALARA principle, should be implemented. Of major interest is estimation the effective dose which operators will receive during a year, considering all operations in that he is involved and all the different possible paths of irradiation or contamination (inhalation, skin penetration, injury, etc.). The estimation of doses received by operating personnel will take into consideration the following steps: - the determination of jobs and events which could involve a significant radiation dose exposure; - whole body and extremities exposure doses should be assessed taking into consideration that the likelihood of contact with radiation and/or contamination sources is higher for hands and legs; - all possible paths of exposure will be identified (external irradiation is the most expected while the internal exposure due to intake could happen following an accidental inhalation of radionuclides or an injury in contaminated medium); - technological controls and administrative measures for exposure minimization will be rigorously implemented; - estimated doses will be compared with maximum permissible levels. The paper describes some general methodologies for computing the total effective doses received by workers involved in decommissioning operations, as well as their application for few special situations, that could contribute significantly to

Study of the WWR-S IFIN-HH reactor main components stare, after 40 years working, using nondestructive methods

International Nuclear Information System (INIS)

Dragolici, A. C.; Zorliu, A.; Ripeanu, R.; Petran, C.; Mincu, I.

2000-01-01

The main goal of these investigations was to establish the security level after 40 years of working of the WWR-S research reactor of Horia Hulubei National Institute of Research and Development for Physics and Nuclear Engineering, Bucharest-Magurele. The purpose of these investigations was: checking the functionality and the physical integrity of the main components of the reactor. The physical integrity of the components is usually affected by slow processes, such as: corrosion, erosion, aging, deformations and initially hidden flaws with very slow evolutions. The methods used to determine the effects of these processes and to infer conclusions about the physical integrity of the facility are: visualizations by optical means (endoscopy and video camera), examination using ultrasounds and gammagraphy. The objective of the endoscopic checking was the view of the state of interior surfaces of the tubes and pipes, specially the inaccessible areas of the non-dismantling parts of the reactor. Big size components, such as reactor vessel, the biologic protection vessel and the main large diameter pipes of the primary cooling system, were investigated using a special device that contains a video camera connected to a PC. To obtain more information regarding the evolution of the corrosion spots, scratches and harmed areas on the investigated surfaces, their depth was checked by ultrasounds, and the welding seams structure was determined by gammagraphy. A table is given with some significant results obtained from ultrasound measurements in different points of reactor vessel, thermal column, horizontal tubes, etc. After these tests, the conclusions are: the maximum corrosion depth is 0.2 mm; - scratches are superficially, not exceeding 0.2-0.5 mm; - the traces of harmed areas are produced by the electromagnetic device utilization used for manipulation of aluminium capsules which contain irradiated substances. They are superficial, with maximum area of about 1 cm 2 ; the

Accuracy of WWR-M criticality calculations with code MCU-RFFI

International Nuclear Information System (INIS)

Petrov, Yu.V.; Erykalov, A.N.; Onegin, M.S.

1999-01-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about Δρ≅±0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, α) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Accuracy of WWR-M criticality calculations with code MCU-RFFI

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation); Erykalov, A N; Onegin, M S [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation)

1999-10-01

The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about {delta}{rho}{approx_equal}{+-}0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, {alpha}) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

Power reactor noise measurements in Hungary

International Nuclear Information System (INIS)

Pallagi, D.; Horanyi, S.; Hargitai, T.

1975-01-01

An outline is given of the history of reactor noise research in Hungary. A brief description is given of studies in the WWR-SM reactor, a modified version of the original WWR-S thermal reactor, for the detection of in-core simulated boiling by analysis of the noise of out-of-core ionization chambers. Coolant velocity measurements by transit time analysis of temperature fluctuations are described. (U.K.)

Seismic safety review mission Almaty WWR 10 MW research reactor Almaty, Kazakhstan. Final report

International Nuclear Information System (INIS)

Gurpinar, A.; Slemmons, D.B.; David, M.; Masopust, R.

1995-06-01

On the request of the government of Kazakhstan and within the scope of the TC project KAZ/0/004, a seismic safety review mission was conducted in Almaty, 8-19 May 1995 for the WWR 10 Mw research reactor. This review followed the fact finding mission which visited Almaty in November 1993 together with an INSARR mission. At that time some information regarding the seismotectonic setting of the site as well as the seismic capacity of the facility was obtained. This document presents the results of further work carried out on both the issues. It discusses technical session findings on geology, seismology, structures and equipments. In the end conclusions and recommendations of the mission are given. 4 refs, figs, tabs, 18 photos

Seismic examination for assessment of safety of location of atomic energy objects (by the example of the WWR-K reactor, Ala-Tau village)

International Nuclear Information System (INIS)

Belyashova, N.N.

2001-01-01

In the Republic of Kazakhstan there are 3 research reactors (the fourth one is temporarily stopped). One of the reactors in 1998 (WWR-K, situated in the Ala Tau village, nearby Almaty city) was conserved because of a number of reasons. Including the reason of the earth crust geological structure insufficient study for the ensuring the seismic safety of the reactor site location. In 1994-1996 a number of geological-geophysical studies was carried out by Kazakhstan specialists confirming the the geological-geophysical conditions in the reactor site location in view of its safety. These condition are meeting to IAEA requirements and up-to-date standards acting in Kazakhstan

Experiment on search for neutron-antineutron oscillations using a projected UCN source at the WWR-M reactor

Science.gov (United States)

Fomin, A. K.; Serebrov, A. P.; Zherebtsov, O. M.; Leonova, E. N.; Chaikovskii, M. E.

2017-01-01

We propose an experiment on search for neutron-antineutron oscillations based on the storage of ultracold neutrons (UCN) in a material trap. The sensitivity of the experiment mostly depends on the trap size and the amount of UCN in it. In Petersburg Nuclear Physics Institute (PNPI) a high-intensity UCN source is projected at the WWR-M reactor, which must provide UCN density 2-3 orders of magnitude higher than existing sources. The results of simulations of the designed experimental scheme show that the sensitivity can be increased by ˜ 10-40 times compared to sensitivity of previous experiment depending on the model of neutron reflection from walls.

Decommissioning of the research nuclear reactor WWR-S Magurele - Bucharest. General presentation of the project

International Nuclear Information System (INIS)

Dragulescu, Emilian; Dragusin, Mitica; Popa, Victor; Boicu, Alin; Tuca, Carmen; Iorga, Ioan; Vrabie, Ionut; Mustata, Carmen

2003-01-01

A decommissioning project was worked out concerning the nuclear facility research reactor WWR-S Magurele-Bucharest to remove the radioactive and hazardous materials and so to exclude any risk for human health and environment. The project involves the four phases named assessment, development, operations and closeout. There are two major parts to the assesment phase: preliminary characterisation and the review and decision-making process. Characterisation is needed to develop project baseline data, which should include sufficient chemical, physical, and radiological characterisation to meet planning needs. Based on the conclusions of these studies, possible decommissioning alternative will be analyzed and: the best alternative chosen, final goal identified, risk assessments are evaluated. Also, taken into account are: regulations supporting assessment, land use considerations, financial concerns, disposal availability, public involvement, technology developments. After a decommissioning alternative was chosen, detailed engineering will begin following appropriate regulatory guidance. The plan will include characterisation information, namely: review of decommissioning alternatives; justification for the selected alternative; provision for regulatory compliance; predictions of personnel exposure, radioactive waste volume, and cost. Other activities are: scheduling, preparation for decommissioning operations; coordination, documentation, characterization report, feasibility studies, Decommissioning Plan, project daily report, radiological survey, airborne sampling records, termination survey of the site. The operations imply: identification and sequencing the operations on contaminated materials, storing on site the wastes, awaiting processing or disposal, and packaging of materials for transport to processing or disposal facilities.The key operations are: worker protection, health and safety program, review of planing work, work area assessment, work area controls

Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

International Nuclear Information System (INIS)

Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

2000-01-01

U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

Waste generated by the future decommissioning of the Magurele VVR-S Research Reactor

International Nuclear Information System (INIS)

Dragolici, F.; Turcanu, C.N.; Dragolici, A.C.

2001-01-01

Nuclear Research Reactor WWR-S from the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei', Bucharest-Magurele, was commissioned in July 1957 and it was shut down in December 1997. At the moment the reactor is in conservation state. During its operation this reactor worked at an average power of 2MW, almost 3216 h/year, producing a total thermal power of 230 x 10 3 MWh. No major modifications or improvements were made during the 40 years of operation to the essential parts of the reactor, respective to the primary cooling system, reactor vessel, active core and electronic devices. So, all components of the measure, control and protection systems are old, generally at the technical level of the 1950s, therefore a reason why in December 1997 the operation was ceased. At present, the reactor can be considered, by IAEA definition in the first stage (reactor shut down, but the vital functions are maintained and monitored). The survey is related to the second stage - restrictive use of the area. To develop a real decommissioning project, it was first necessary to evaluate the volume and the characteristics of the radioactive waste which will be generated. Radioactive waste generated during the decommissioning of Magurele WR-S research reactor may be classified as: Activated wastes (internal structures, horizontal channels and thermal column, biological shield); Contaminated wastes (primary circuit non-activated components, hot cells, some technological rooms as main hall, pumps room, radioactive material transfer areas, ventilation building and stack); Possibly contaminated materials from any area of reactor building and ventilation building. After 40 years of nuclear research activities, all such areas are suspected of contamination. The volume of wastes that will result from WWR-S Research Reactor decommissioning is summarized

Generation of the problem-dependent data libraries for IFIN-HH WWR-S spent fuel storage criticality and dose calculation

International Nuclear Information System (INIS)

Ene, Daniela; Tigau, F.

1998-01-01

The methods used for the radioactivity inventory calculation and dose evaluation of the fuel elements irradiated in the WWR-S IFIN-HH reactor are discussed in this work. A particular attention is paid to the processed problem-dependent nuclear libraries. SAS2H, a complex sequence of the SCALE-4.3 code system containing the modules BONAMI - NITAWL - XSDRNPM - COUPLE - ORIGEN-S - XSDOSE, has been assimilated on the IFIN-HH computer and applied to update the ORIGEN-S libraries by producing problem-dependent processed data libraries needed to perform the depletion and shielding analysis. This sequence uses one of the eight associated data libraries of the SCALE-4.3 system according to the choice of the user. The method consists in the following analysis processes: i) lattice cell neutron analysis to produce the flux weighting spectrum for activation library updating; ii) update of the nuclear data constants of the ORIGEN-S libraries; iii) depletion and decay analysis for a specified fuel assembly and irradiation history in order to generate gamma and neutron source strength and spectra. iv) one-dimensional radial shielding calculation for the evaluation of the angular neutron and gamma flux at the surface of a spent fuel shipping cask and further calculation of the dose rates at various points outside the cask. An efficient alternative of the calculation sequence mentioned above is the ARP (Automatic Rapid Processing) method conceived in order to generate independently ORIGEN-S libraries and to reduce substantially the running time. The substance of this method is the generation of the problem-dependent libraries from basis libraries a priori created by SAS2H for specific fuel assembly type and further interpolation of two independent variables, enrichment and burnup. Specific applications concerning WWR-S spent fuel were performed: i) generation of three problem-dependent libraries for the S-36 fuel assembly taking into account the maximum value of the burnup of this

Calculational investigations and analysis of characteristics of research reactor WWR-M as a source of neutrons for solution of scientific and applied tasks

International Nuclear Information System (INIS)

Vorona, P.M.; Razbudej, V.F.

2010-01-01

Calculational studies and analysis of the neutron fields of WWR-M research reactor of the Institute for Nuclear Research, National Academy of Sciences of Ukraine, as a basic nuclear facility for performing the fundamental and applied investigations and for experimentalindustrial production of radioisotope products for various spheres of application are carried out. The calculations are carried out by the method of statistic tests (Monte Carlo) applying the computer program MCNP-4C. The data on the spectra and the neutron flux density values at the 10 MW reactor power for all technological facilities designed for the works with neutrons: 19 vertical experimental channels for irradiation of specimens and 10 horizontal channels for beams extraction from the reactor are obtained. The effect of the neutron traps (water cavities) mounted in the core on the characteristics of the extracted from the reactor beams is demonstrated. Recommendations associated with optimization of the reactor core are adduced for amplification of its capabilities as a neutron source in experimental researches.

Encapsulation technology of MR6 spent fuel and quality analysis of the EK-10 and WWR-SM spent fuel stored more than 30 years in wet conditions

Energy Technology Data Exchange (ETDEWEB)

Borek-Kruszewska, E.; Bykowski, W.; Chwaszczewski, S.; Czajkowski, W.; Madry, M. [Institute of Atomic Energy, Otwock -Swierk (Poland)

2002-07-01

The research reactor MARIA has been in operation for more than twenty years and all the spent fuel assemblies used since the first commissioning of the reactor are stored in wet facility on site. The present paper deals with the spent fuel MR-6 encapsulation technology in MARIA reactor. The encapsulated spent MR-6 fuel will be stored under water in the same pool unless some other solution is available. The capsules made of stainless steel are capable to accommodate one MR-6 fuel assembly. The encapsulation process is performed in the hot cell by the MARIA reactor. The spent fuel having its leg cut off is loaded to the transport cylinder manually and next transferred to a trolley. The trolley is moving to a position directly below the entrance to the hot cell and the spent fuel is entering the hot cell. The spent fuel assembly is then put into the drying cell. Dried out spent fuel is moved into the capsule mounted on the grip of the machine. Next, the capsule lid is pressed in and welded. After the leak test and filling up with helium the capsule returns from the hot cell to the pool. The hermetic capsule is sunk back into the water and positioned in the separator . The results presented earlier show, that the limiting time of WWR-SM and Ek-10 type spent fuel residence in wet storage is about 40-45 years. Therefore, the systematic quality investigation of all Ek-10 fuel elements and WWR-SM fuel assemblies discharged from EWA reactor in the period of 1959-1969 was performed. Altogether, about 2500 Ek-10 fuel elements and 47 WWR-SM fuel assemblies were investigated. The results of these investigations are presented in the present work. The sipping test, visual investigation and ultrasonic techniques were used for that purpose. The radioactive isotope Cs-137 was used as the indicator of fission product release from the fuel assembly. Taking into account the value of Cs-137 release from damaged WWR-SM fuel assembly the criteria of damaged fuel assembly were proposed. It

Radiologic states of the WWR-S Bucharest Reactor following definitive shutdown

International Nuclear Information System (INIS)

Garlea, C.; Kelerman, C.; Mocioiu, D.; Garlea, I.

2001-01-01

The definitive shutdown of a reactor raises problems related to the management of the radioactive inventory. To define the radioactive inventory contained in the burned nuclear fuel and in the neutron activated structural materials computation methods are to be used. Besides the radioactive inventory contained in the main block of the reactor, the one due to the primary circuit contaminated mainly with fission products and corrosion products activated in the reactor core, transported and deposed on the components of the cooling primary circuit should be added. Also another component of the radioactive inventory intervenes, namely, the one due to the contamination of the technological rooms used for various operations the nuclear activities (hot cells, pump room, reactor hall, passage ways to the hot cells and for radioactive source, radioisotope and radioactive waste transport). The activities which made used of the neutron and gamma fluxes for radioisotope production, materials irradiation, research, component testing, resulted in radioactive waste, technological or accidental contaminations of the technological rooms of the reactor. Inspections and current repair interventions resulted also in radioactive waste an contaminations. Consequently systematic measurements with qualified equipment dedicated to alpha, beta, gamma contamination measurements as well as to dose rates determinations for the personnel exposed are necessary. Irrespective of the duration of the reactor conservation or shutdown, the radiologic monitoring should continue. This work presents the results obtained by the research group 'Restoration of Nuclear Sites', working with the IFIN-HH, regarding both the radioactive inventory calculation and measurements of contamination of technological rooms and environment in the reactor vicinity

Supply of Prague with heat from a nuclear heat source

International Nuclear Information System (INIS)

Poul, F.

1976-01-01

The proposals are discussed of supplying Prague, the Czechoslovak Capital, with nuclear reactor-generated heat energy. The proposals meet the requirements of the general urban plan of development. The first nuclear heating plant is to be sited in the Kojetice locality, in the northern Prague suburb. It will be commissioned by 1984 and 1985. It is estimated that the maximum heat output in form of hot water will be 821 MW. By 1995 the construction of the second nuclear heating plant should be started southeast or east of Prague. The connection of these two nuclear plants to the hot water mains together with other conventional heating plants will secure the heat supply for Prague and its new housing estates and industrial works. (Oy)

Calculation of prefabricated part of WWR-K reactor building

International Nuclear Information System (INIS)

Belyashova, N.N.; Aptikaev, F.F.; Kopnichev, Yu.F.

1998-01-01

According of factual characteristics a strength and deformation of over-land part of carrier constructions under construction movement is defined. Direct dynamical calculation of design elements under action of inertial loads from supports shifts shows, that seismic stability of enclosing construction is not ensured. Possibly practically total collapse of coating construction is possibly, under which following levels of damages of internal design constructions of reactor central room have been forecasted: 1. Fall of destroyed design construction on reactor vessel in time moment (1.56-1.59 s) after coming to building of earthquake seismic waves of 10 balls. 2. It is possibly cracks formation in radial direction in lower part of reactor cap, but destroying of cap does not incident; 3. It is possibly cracks formation within stretched concrete zone of reactor construction at the mark from - 0.859 up to 0.100. Destroy of concrete's compressive zone of reactor construction have not being expected. 4. Collapse of reactor first contour coating constructions have not being expected

Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

International Nuclear Information System (INIS)

Romanova, N.K.; Takemoto, N.

2013-01-01

Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

Andalusi sherds from Prague

Czech Academy of Sciences Publication Activity Database

Charvát, Petr; Hrdlička, Ladislav; Delery, C.

2013-01-01

Roč. 65, č. 1 (2013), s. 198-206 ISSN 0323-1267 Institutional support: RVO:67985912 Keywords : Middle Ages * Prague * commerce * Spain * Almohad empire Subject RIV: AC - Archeology, Anthropology, Ethnology

Two semidetached single family houses in Prague

OpenAIRE

DIEGO JOVELLS, FERNANDO JOSÉ

2015-01-01

Trabajo Fin de Grado de modalidad Intercambios Académicos. České Vysoké Učení Technické v Praze | Czech Technical University in Prague [en] This Bachelor’s Thesis is intent on building a single family house in Veleslavín, a district of Prague, part of Prague 6, situated in the west of the city. The chosen foundations, the structure, the installations of electricity, cold and hot water, heating system, and sanitation network are going to be studied in this project. Some plans will be done, ...

Geochemical characteristics of black shales from the ore-bearing complex of strata of the Male Karpaty Mts. IV. Evaluation of data on the element contents obtained by neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Cambel, B. (Slovenska Akademia Vied, Bratislava (Czechoslovakia). Geologicky Ustav); Khun, M. (Komenskeho Univ., Bratislava (Czechoslovakia). Prirodovedecka Fakulta)

1983-06-01

The application is described of instrumental neutron activation analysis in determining the levels of the individual elements in rocks. Samples were irradiated in the WWR-S reactor at Rez near Prague, and evaluated by the Radiometric Department of the Central Laboratories of the Czechoslovak Uranium Industry. The results are described in detail of measuring the levels of W, Ta, Sc, Hf, Ga, Rb, Cs, U, Th, Au, Sb, As, Zn and Ag in Male Karpaty rocks. The levels of the elements in different samples and in different localities are tabulated. A comparison is made with the Keno Hill, Yukon (USA) area where the polymetallic mineralization is evolutionally analogous to that of the Male Karpaty region.

Geochemical characteristics of black shales from the ore-bearing complex of strata of the Male Karpaty Mts

International Nuclear Information System (INIS)

Cambel, B.; Khun, M.

1983-01-01

The application is described of instrumental neutron activation analysis in determining the levels of the individual elements in rocks. Samples were irradiated in the WWR-S reactor at Rez near Prague, and evaluated by the Radiometric Department of the Central Laboratories of the Czechoslovak Uranium Industry. The results are described in detail of measuring the levels of W, Ta, Sc, Hf, Ga, Rb, Cs, U, Th, Au, Sb, As, Zn and Ag in Male Karpaty rocks. The levels of the elements in different samples and in different localities are tabulated. A comparison is made with the Keno Hill, Yukon (USA) area where the polymetallic mineralization is evolutionally analogous to that of the Male Karpaty region. (M.D.)

Determination of the gamma-ray flux of the stopped WWR-SM reactor by color center production in LiF

International Nuclear Information System (INIS)

Mussaeva, M.A.; Kalannov, M.U.; Ibragimova, E.M.; Karabaev, Kh.Kh.

2004-01-01

Full text: Gamma-radiation with a wide energy spectrum, accompanying neutron flux in the nuclear reactor, is known to result in radiation heating of materials. It is usually detected either by calorimetry or by an ionizing chamber maintained in the active zone while the reactor works and high-energy neutrons also contribute into ionization. The aim of this research was to separate the gamma-component from the neutron flux upon stopping the WWR-SM reactor and to determine the gamma-intensity both with the ionization chamber and the well-known dosimeter LiF crystal, and also by comparing with the effect of monochromatic 60 Co gamma-radiation of the known flux and dose. For LiF with small Z the photoelectric effect is weak, and Compton scattering prevails. Both the optical absorption and photo-luminescence techniques together with micro-hardness and X-ray diffraction analysis were used for measuring the structure defect generation rate in the irradiated crystals, which is proportional to the gamma-intensity. Fluorine vacancy trapping electron is the well-known stable F-center responsible for the isolated absorption band at 250 nm and induced by radiolysis mechanism. The sequential irradiations and measurements were done within 150 hours after the moment of the reactor quenching. The dose dependence of the absorption band was found to be linear up to the dose of 10 6 R. The F-center concentration as a measure of an accumulated dose was calculated by the Smakula formula. At higher doses another band at 440 nm appears like that for 60 Co irradiation, which is responsible for unstable F 2 and F 3 centers formed due to coagulation of F-centers. X-diffraction analysis revealed twin structure in (111) plane. Yet the micro-hardness of the gamma-irradiated samples did not change noticeably. For higher doses the photo-luminescence band at 650 nm was also used as a dosimetric item. The luminescence kinetics has a fast nanosecond scale component and a weak tail in a microsecond

“Intonazioni degli Ebrei” from Benedetto Marcello’s Estro poetico-armonico in Prague in 1729

Czech Academy of Sciences Publication Activity Database

Jonášová, Milada

2015-01-01

Roč. 52, č. 1 (2015), s. 5-54 ISSN 0018-7003 R&D Projects: GA ČR(CZ) GAP409/12/2563 Keywords : Benedetto Marcello * Intonazioni degli Ebrei * Prague Subject RIV: AL - Art, Architecture, Cultural Heritage

[Radiation ecological environment in the Republic of Kazakhstan in the vicinity of the reactors and on the territory of the Semipalatinsk Test Site].

Science.gov (United States)

Kim, D S

2012-01-01

The results of research into the environmental conditions in the regions of location of the pressurized water reactor WWR-K, fast neutron breeder BN-350 and on the territory of the Semipalatinsk Test Site are represented. The effects of the exposure to aerosol emissions from WWR-K and BN-350 reactors on the environment are summarized. We present some arguments in favor of the safe operation of fission reactors in compliance with the rules and norms of nuclear and radiation protection and the efficient disposal of radioactive waste on the territory of the Republic.

Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

International Nuclear Information System (INIS)

Hanan, N.A.; Liaw, J.R.; Matos, J.E.

2005-01-01

A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

The Conference from the Prague Perspective

Czech Academy of Sciences Publication Activity Database

Hrubec, Marek

2017-01-01

Roč. 43, č. 3 (2017), s. 256-257 ISSN 0191-4537 Institutional support: RVO:67985955 Keywords : Critical Theory * Conference Philosophy and Social Science * Prague Subject RIV: AA - Philosophy ; Religion

Computerized reactor power regulation with logarithmic controller

International Nuclear Information System (INIS)

Gossanyi, A.; Vegh, E.

1982-11-01

A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

National Gallery in Prague

Czech Academy of Sciences Publication Activity Database

Mikeš, Stanislav; Haindl, Michal; Holub, Radek

2011-01-01

Roč. 2011, č. 86 (2011), s. 23-24 ISSN 0926-4981 R&D Projects: GA MŠk(CZ) LG11009 Institutional research plan: CEZ:AV0Z10750506 Keywords : virtrual reality * information system Subject RIV: BD - Theory of Information http://library.utia.cas.cz/separaty/2011/RO/haindl-national gallery in prague.pdf

The Rossendorf research reactor. Operating and dismantling from a point of view of the emission control; Der Rossendorfer Forschungsreaktor. Betrieb und Rueckbau aus Sicht der Emissionsueberwachung

Energy Technology Data Exchange (ETDEWEB)

Bauer, B.; Beutmann, A.; Kaden, M.; Scheibke, J. [VKTA, Dresden (Germany); Boessert, W.; Jansen, K.; Walter, M.

2016-07-01

The Rossendorf research reactor went in operation in 1957 as GDR's first nuclear reactor and Germanys second after FRM Garching. It was a heterogeneously structured, light-water moderated and cooled tank-reactor of the Soviet type WWR-S. During his time of operation, he served both the research and the production of radioisotopes. The history of exhaust air emission monitoring and its results are presented. With view to the decommissioning time selected results are discussed. The estimated discharges are compared by the actually recognized.

Einstein and Prague

International Nuclear Information System (INIS)

Bicak, J.

1979-01-01

A commemorative publication is submitted issued on the occasion of Albert Einstein's centenary remembering Einstein's sojourn and work in Prague. In addition to the article Ueber den Einfluss der Schwerkraft auf die Ausbreitung des Lichtes the publications contains the author's preface to the Czech edition of his Theory, the list of studies written by Einstein while in Prague, an assay on the great physicist's life and work, and extracts from the memoires of Philipp Frank published under the title Einstein, His Life and Times. (L.O.)

The application of calorimetrical methods in nuclear technology and dosimetry

International Nuclear Information System (INIS)

Kott, J.; Krett, V.; Novotny, J.; Kovar, Z.; Jirousek, V.

1985-01-01

The report reviews theoretical as well as experimental research activities devoted to the possibilities of measuring reactor neutron and photon fields using thermic detectors based on calorimetric principle. There have been worked out theoretical principles of a reactor measuring probe intended in the first place to measuring neutron fluxes under operational temperatures inside power and research reactors, and a new philosophy of measurement has been elaborated. In addition, the report presents the experimental results as obtained on research reactors WWR-S, WWR-SM, RA, and Czechoslovak power reactor A-1 and GDR power reactor WWR-2. These results are given in connection with a newly proposed technique of reactor neutron field detection. The second part of the report presents results of works concerning beam dosimetry with the use of calorimeters

The Prague Linguistic Circle and Dialectics

Czech Academy of Sciences Publication Activity Database

Sládek, Ondřej

-, č. 19 (2017), s. 352-357 E-ISSN 2037-2426 Institutional support: RVO:68378068 Keywords : The Prague Linguistic Circle * Jan Mukařovský * Structuralism * Structural Poetics * Dialectics Subject RIV: AJ - Letters, Mass-media, Audiovision OBOR OECD: Specific literatures

PROMOTION AND MARKETING OF BUCHAREST - PRAGUE TOURS

Directory of Open Access Journals (Sweden)

Petrica STEFAN

2014-10-01

Full Text Available This paper has as its theme the marketing and promotion of tourism circuit between Bucharest and Prague, unique area in the world because there is a harmonious combination of potential natural and cultural values and historical tourism which attracts many tourists. Also due to the geographical potential tourists coming to Prague can practice sports activities both winter and summer.Both natural resources and the human touch are important sights to attract more tourists in the area, and in Prague, the potential exists.

Safety operation of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.

2001-01-01

There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

Irradiation facilities for the production of radioisotopes for medical purposes and for industry at the Rossendorf Research Reactor

International Nuclear Information System (INIS)

Hieronymus, W.

2007-01-01

In 1955, the Government of the German Democratic Republic initiated radioisotope production. With that decision, the following plants received their go ahead: - Research reactor with its user facilities; - Cyclotron with its specific facilities; - Institute for radiochemistry; - Library, lecture hall, workshops and administration buildings supporting the necessary scientific and administrative environment. The Zentralinstitut fuer Kerntechnik (ZfK), also known as the Central Institute for Nuclear Technology, was founded at Rossendorf near Dresden, Germany, to house all those plants. The Rossendorf Research Reactor (RFR) was constructed in 1956-1957. That endeavour was enabled by the technological support of the former USSR under a bilateral agreement which included the delivery of a 2 MW research reactor of the WWR-S design

Preparing the construction of a school reactor

International Nuclear Information System (INIS)

Matejka, K.

1977-01-01

The possibilities are discussed of teaching and training nuclear reactor operation and control, teaching experimental reactor physics and investigating reactor lattice parameters using a training reactor to be installed at the Faculty of Nuclear Science and Physical Engineering in Prague. Requirements are indicated for the reactor's technical design and the Faculty's possibilities to contribute to its construction. (J.B.)

Relativity and Gravitation : 100 Years After Einstein in Prague

CERN Document Server

Ledvinka, Tomáš; General Relativity, Cosmology and Astrophysics : Perspectives 100 Years After Einstein's Stay in Prague

2014-01-01

In early April 1911 Albert Einstein arrived in Prague to become full professor of theoretical physics at the German part of Charles University. It was there, for the first time, that he concentrated primarily on the problem of gravitation. Before he left Prague in July 1912 he had submitted the paper “Relativität und Gravitation: Erwiderung auf eine Bemerkung von M. Abraham” in which he remarkably anticipated what a future theory of gravity should look like. At the occasion of the Einstein-in-Prague centenary an international meeting was organized under a title inspired by Einstein's last paper from the Prague period: "Relativity and Gravitation, 100 Years after Einstein in Prague". The main topics of the conference included: classical relativity, numerical relativity, relativistic astrophysics and cosmology, quantum gravity, experimental aspects of gravitation, and conceptual and historical issues. The conference attracted over 200 scientists from 31 countries, among them a number of leading experts in ...

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

Full text: The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by (1) implementing safety upgrades, or (2) assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

International Nuclear Information System (INIS)

Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

2004-01-01

The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by implementing safety upgrades, or assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

Marketingová komunikace a efektivita reklamní kampaně Nike Run Prague 2011 (sémiotická a narativní analýza)

OpenAIRE

Housková, Jitka

2012-01-01

The thesis deals with the analysis of the marketing communication and advertising campaign of the running event Nike Run Prague, which took place in Prague 3 September 2011. It also deals with semiotic and narrative means and with the research of the effectiveness of the campaign. Nike Run Prague is a 10 km long run throughout Prague and it was organized by Nike for the very first time. The running event was amply publicized and within its advertising campaign adopted the marketing strategy t...

Creation of reactor's reliable system of emergency energy supply

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Brovkin, A.Yu.; Petukhov, V.K.; Chekushin, A.I.; Chernyaev, V.P.; Yagotinets, N.A.

1998-01-01

System of reliable power supply of the WWR-K reactor complex is described, which completely provides safety operation of reactor equipment in the case of total voltage loss from external power transmission lines as well as under destruction of accumulation batteries by earthquake more than 6 balls. Switching on in operation of diesel-generators and system of constant current supply from accumulator batteries is occurred automatically under cessation of voltage supply from centralized power system. Reliable reactor dampening in case it work on capacity has been ensured. Reactor cooling under its emergency shutdown during both the partial or the total loss of coolant in first counter has been carried out. Under full coolant loss the system of emergency reactor cooling has been switched on in operation

Implementace marketingového mixu Prague City Golf Club

OpenAIRE

Čmakal, Tomáš

2012-01-01

Title: Marketing mix implementation of Prague City Golf Club Objective: Analysis of marketing mix of the golf resort Prague City Golf Club and addresses proposals for its further improvement Methods: PEST analysis SWOT analysis Porter five forces analysis In-depth interview Results: The results of assessment of the marketing mix instruments of the Prague City Golf Club have shown that it is a high-quality golf resort with elaborate and numerous components of the marketing mix. In terms of com...

Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

International Nuclear Information System (INIS)

Arkhangelsky, N.

2002-01-01

The Russian RERTR Program started at the end of 70's, the final goal of the program is to eliminate supplies of HEU in fuel elements and assemblies for foreign research reactors that were designed according to Russian projects. Basic directions of the work include: completion of the development of the fuel elements and assemblies on a basis of uranium dioxide; development of the fuel on a basis of U-Mo alloy; and development of pin type fuel elements. Fuel assemblies of WWR-M2 type with LEU were developed and qualified for using in foreign research reactors that use such type of fuel assemblies. These assemblies are ready for the supplying several operating foreign research reactors. There are more than 20 sites in Eastern European countries, former Soviet republics and another countries that have big amount of Russian origin HEU in fresh and spent fuel. The problem of the shipment of SNF from sites of research reactors is also very important for domestic Russian research reactors. More than ten years from its beginning the Russian RERTR program developed practically independently from the international RERTR program and only at the begin of 90's the Russian specialists started to contact with foreign scientists and the exchange of the scientific information has become more intensive. In September 1994, representatives of Minatom and DOE signed a protocol of intent to reduce an enrichment of uranium in research reactors. The main aspects of collaboration involve: Several domestic Russian research reactors such as WWR-M, IR-8 and others were investigated from the point of view of possibility of reducing of enrichment; financial support of the program from US DOE which is insufficient. The important part of international collaboration is the import of Russian origin spent and fresh fuel of research reactors to Russia. In August 2002 an impressive result of the Russian-American collaboration with support of IAEA and with the help and assistance of Yugoslavian side was

Trends in precipitation variability: Prague (the Czech Republic)

Czech Academy of Sciences Publication Activity Database

Bodri, L.; Čermák, Vladimír; Krešl, Milan

2005-01-01

Roč. 72, č. 1-2 (2005), s. 151-170 ISSN 0165-0009 R&D Projects: GA AV ČR IAA3012005; GA ČR GA205/03/0998 Institutional research plan: CEZ:AV0Z30120515 Keywords : precipitation variability * Prague observatory * North Atlantic Oscillation Subject RIV: DC - Siesmology, Volcanology, Earth Structure Impact factor: 2.479, year: 2005

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

International Nuclear Information System (INIS)

Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

2004-01-01

Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

La figure de l'autorité magistrale à travers Jan Hus et Jérôme de Prague

Czech Academy of Sciences Publication Activity Database

Pavlíček, Ota

2011-01-01

Roč. 85, č. 3 (2011), s. 371-389 ISSN 0035-2217 R&D Projects: GA AV ČR(CZ) KJB900090903 Institutional research plan: CEZ:AV0Z90090514 Keywords : Jerome of Prague * John Hus * magisterial authority * truth * medieval thought * nationalism * Prague University Subject RIV: AA - Philosophy ; Religion

Analýza strategie hotelu Hilton Prague

OpenAIRE

Lišková, Táňa

2011-01-01

The goal of this thesis is to analyze the strategy of Hotel Hilton Prague. After introduction of the company there is a brief description of marketing and some specification regarding the service and hotel marketing. In following part I focus on the PEST analysis and SWOT analysis of Hilton Prague. Next chapter is regarding the marketing mix of the hotel, giving a detailed description of all the components.This chapter is followed by market segmentation, customer and competition analysis. Cus...

St. George at Prague Castle and Perseus: an Impossible Encounter?

Czech Academy of Sciences Publication Activity Database

Bažant, Jan

2015-01-01

Roč. 19, 1/2 (2015), s. 189-201 ISSN 1212-5865 Institutional support: RVO:67985955 Keywords : Bellerophon * St. George * iconography * Prague Castle * classical tradition Subject RIV: AB - History http://studiahercynia.ff.cuni.cz/wp-content/uploads/sites/79/2016/05/jan_bazant_189-201.pdf

Status of Dalat research reactor and progress of new reactor plan in Vietnam

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Vien, Luong Ba

2005-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

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)

Experience in reactor research and development programs as educational system for thermohydraulic engineering

International Nuclear Information System (INIS)

Zaki, G.M.; Fikry, M.M.

1977-01-01

A reactor development program within a research reactor facility can be used for personnel training on the operation of power reactors and research in the different fields of nuclear science and engineering. A training program is proposed where reactor maintenance and operation, in addition to conducting development programs and executing projects, are utilized for forming specialized groups. The paper gives a short survey of a heat transfer program where out of pile and in-core studies are conducted along with two-phase flow investigations. This program covers the main requirements for WWR (water cooled and moderated reactor) power uprating and furnishes basic knowledge on power reactor thermal parameters. The major facilities for conducting similar programs devoted to education are mentioned

Estimation of the outlooks for large-scale transmutation of fission-produced iodine

CERN Document Server

Galkin, B Y; Kolyadin, A B; Kocherov, N P; Lyubtsev, R I; Hosov, A A; Rimskij-Korsakov, A A

2002-01-01

To obtain data necessary for estimating sup 1 sup 2 sup 9 I transmutation efficiency in nuclear reactors the effective neutron capture cross section on sup 1 sup 2 sup 9 I isotope in neutral spectrum of the WWR-M reactor was determined. The calculated value of sup 1 sup 2 sup 9 I capture cross section, averaged by neutron spectrum in beryllium reflector of the WWR-M reactor, made up 17.8+-3.2 barn. On the basis of experimental data and estimations it was shown that in neutron flux 10 sup 1 sup 4 1/(cm sup 2 s) transmutation of iodine -129 loaded in the course of one year can amount to approximately 25%

Photochemical Degradation of Polybrominated Diphenyl Ethers in Micro Photo-Reactor

Czech Academy of Sciences Publication Activity Database

Vajglová, Zuzana; Veselý, M.; Křišťál, Jiří; Vychodilová, Hana; Tříska, Jan; Jiřičný, Vladimír

2012-01-01

Roč. 42, SI (2012), s. 1378-1382 E-ISSN 1877-7058. [International Congress of Chemical and Process Engineering CHISA 2012 and 15th Conference PRES 2012 /20./. Prague, 25.08.2012-29.08.2012] R&D Projects: GA ČR GA104/09/0880 Institutional support: RVO:67985858 ; RVO:67179843 Keywords : polybrominated diphenyl ethers * micro photo reactor * debromination Subject RIV: CI - Industrial Chemistry, Chemical Engineering

'Ipsa dicit, quod sic est, ergo verum.' Authority of Scripture, the Use and Sources of Biblical Citations in the Work of Jerome of Prague

Czech Academy of Sciences Publication Activity Database

Pavlíček, Ota

Suppl., č. 2 (2015), s. 70-89 ISSN 0015-1831 Institutional support: RVO:67985955 Keywords : Jerome of Prague * Bible * John Wyclif * Bohemian Reformation * Prague University Subject RIV: AA - Philosophy ; Religion

Ontogenetic and intraspecific variation in the late Emsian - Eifelian (Devonian) conodonts Polygnathus serotinus and P. bultyncki in the Prague Basin (Czech Republic) and Nevada (western U.S.)

Science.gov (United States)

Klapper, Gilbert; Vodrážková, Stanislava

2013-06-01

Klapper, G. and Vodražkova, S. 2013. Ontogenetic and intraspecific variation in the late Emsian - Eifelian (Devonian) conodonts Polygnathus serotinus and P. bultyncki in the Prague Basin (Czech Republic) and Nevada (western U.S.). Acta Geologica Polonica, 63 (2), 153-174, Warszawa. Samples from populations of Polygnathus serotinus Telford 1975 and P. bultyncki Weddige 1977 from the Prague Basin and Nevada display normal variation for Devonian conodont species. A considerable number of previous authors, however, have proposed unnecessary synonyms of these two species, primarily because they have not recognized ontogenetic variation. In contrast, we interpret the variation as ontogenetic as well as intraspecific and present detailed synonymies as a result. A third species, P. praetrigonicus Bardashev 1992, which has been carried in open nomenclature for many years, is an important indicator of the basal costatus Zone in the Prague Basin, New York, and Nevada. We review the stratigraphic distribution of these three species and the conodont zonation across the Emsian-Eifelian (Lower-Middle Devonian) boundary. Polygnathus pseudocostatus sp. nov. (partitus-costatus zones, central Nevada) is described herein. We have observed a decrease in the pit size during ontogeny in P. bultyncki although we have not measured enough specimens to rule out intraspecific versus ontogenetic variation.

Mediální obraz Prague Pride v letech 2011 a 2013

OpenAIRE

Tomašková, Karolína

2015-01-01

The aim of the diploma thesis Media image Prague Pride in 2011 and 2013 is to analyze the representation of LGBT community in the media, define the basic concepts, historical context and media stereotypes associated with it. The carrier backbone of this work is a media image analysis - of the first and third year of Prague Pride festival, which for the queer community meant a crucial milestone. Prague Pride is interesting for more points of view. Although a minor topic, the gay pride has attr...

La théorie du vers et le Cercle linguistique de Prague

Czech Academy of Sciences Publication Activity Database

Ibrahim, Robert; Plecháč, Petr

2014-01-01

Roč. 2, č. 50 (2014), s. 101-114 ISSN 0075-966X R&D Projects: GA ČR GAP406/11/1825 Institutional support: RVO:68378068 Keywords : Prague Linguistic Circle * theory of verse * phonetics and phonology * linguistic * structuralism Subject RIV: AJ - Letters, Mass-media, Audiovision

The feasibility of using a Fourier RTOF spectrometer at a low-power research reactor

International Nuclear Information System (INIS)

Maayouf, R.M.A.; Priesmeyer, H.G.; Kudryashev, V.A.

1991-01-01

The present situation of Fourier time-of-flight (TOF) spectrometry is discussed using the FSS spectrometer as example. The use of the Fourier reverse TOF spectrometry, as an efficient tool for studying condensed matter, at a 2 MW (WWR-S type) reactor is also assessed. The arrangement of the RTOF spectrometer, which could be successfully used at such type of reactor, is introduced. The suggested arrangement applies a neutron guide tube of 24 m length and allows for effective luminosity 2.4.10 6 at a flight path distance of 3.6 m. The number of neutrons scattered from a sample (5 cm 3 in volume) and incident on the detector system, as estimated for the suggested arrangement, is ∝1.6.10 3 n/sec. Such high counting rate allows to measure a diffraction spectrum within less than an hour. (orig.) With 12 figs [de

Podnikatelský plán společnosti AltheaPrague, s.r.o.

OpenAIRE

Štefanová, Tereza

2011-01-01

The aim of this diploma thesis was to create a business plan for newly established private medical company. The first part theoretically deals with business plan as a strategic document. It explains all essential chapters and topics, visual aspects and its purpose. It also deals with the specifics of the private enterprise in health care in the Czech Republic. The second part of the thesis is devoted to practical procedures needed for establishing the centre of preventive medicine AltheaPragu...

The Prague Linguistic Circle and Dialectics

Directory of Open Access Journals (Sweden)

Ondřej Sládek

2017-12-01

Full Text Available The study deals with dialectics in the context of the Prague Linguistic Circle, particularly in the context of Jan Mukařovský’s thinking. The essay presents 1 main sources of Mukařovský’s dialectics, and outlines 2 Mukařovský’s dialectical method. The notion of dialectics appears in Mukařovský’s scholarly work in a set of connections. He applied dialectics as a method, manner or form of rationality. It served as a means of gaining knowledge about the world, specific phenomena and objects, their essence, interconnectedness as well as development. Mukařovský also used it as a procedure for resolving contradictions (antinomies that he encountered in his scientific explorations and in ordinary practical activities. He understood dialectical thinking as dynamic, open, and pluralist thinking striving to reflect reality as a constant process. Gradual coming together of dialectics and materialism, evident in Mukařovský’s scholarly works from the mid-1930s, resulted, ten years later, in a public adoption of dialectical materialism.

Assemblage of wool and silk textiles from medieval waste layers in Prague, Czech Republic

Czech Academy of Sciences Publication Activity Database

Březinová, Helena; Kohout, D.

Roč. 25, č. 6 ( 2017 ), s. 119-125 ISSN 1230-3666 R&D Projects: GA ČR(CZ) GA14-06451S Institutional support: RVO:67985912 Keywords : archaeological textiles * Middle Ages * Prague * waste layers * dyeing analyses Subject RIV: AC - Archeology, Anthropology, Ethnology OBOR OECD: Archaeology Impact factor: 0.626, year: 2016

Requirements for BNCT at a nuclear research reactor. Results from a BNCT workshop organized by the European Commission in Prague, November 2005

International Nuclear Information System (INIS)

Moss, Ray; Sauerwein, Wolfgang; Wittig, Andrea; Burian, Jiri

2006-01-01

As part of the European Commission's Enlargement and Integration Action (E and IA), which is intended to improve exchange and relationship within the extended European Union (EU), a Workshop was organized in Prague in November 2005. The purpose of the workshop was to present and discuss technical and organisational requirements in setting up a BNCT facility at a research reactor. Topics included: treatment of a patient by BNCT; organisational aspects and regulatory affairs; BNCT from the nuclear perspective and BNCT from the clinician's perspective. Presentations were given by BNCT experts in their particular field, whilst eleven different national nuclear research centres from the New Member States and Accession Countries, interested in developing a BNCT programme, presented the status of their preparations. The conclusions of the Workshop were that an early and close collaboration between nuclear and medical groups is the basis for BNCT, that a local effort to build a BNCT facility should be supported by a national research programme including basic and clinical science and that the JRC and its partners are ready to support national initiatives within the EU and candidate countries. (author)

Neutron spectra measurements and neutron flux monitoring for radiation damage purposes

International Nuclear Information System (INIS)

Osmera, B.; Petr, J.; Racek, J.; Rumler, C.; Turzik, Z.; Franc, L.; Holman, M.; Hogel, J.; Kovarik, K.; Marik, P.; Vespalec, R.; Albert, D.; Hansen, V.; Vogel, W.

1979-09-01

Neutron spectra were measured for the TR-0, WWR-S and SR-0 experimental reactors using the recoil proton method, 6 Li spectrometry, scintillation spectrometry and activation detectors in a variety of conditions. Neutron fluence was also measured and calculated. (M.S.)

Aspects of intellectual property related to the TRIGA reactor in Romania

International Nuclear Information System (INIS)

Chirita, Ion

2008-01-01

Full text: A TRIGA - type research reactor has been operating in Pitesti since 1979. In Romania, the first research reactor - of the WWR-C type - has been operating since 1957. Both these reactors have contributed to the formation of well - trained specialists, whose works constitute an important intellectual and industrial property. Institute for Nuclear Research (formerly INT, then INPR) is the holder of several published patents, such as: Procedure for decontamination of water and primary circuits of irradiation devices; Reconditioning of ion exchangers; Nozzle for flow water gaugers; Oscillating electromagnetic pump; Facility for determining nuclear fuel burnup; Portable monitor for contamination measurements; Cable joints with biological protection; Anti-seismic and thermal connection; Automatic facility for nuclear fuel irradiation testing; Method for determining power distribution specific for research rector fuel elements; Tight end-fittings; Cooling damage facility, etc. Many of these have been applied or can be applied to reactors of the TRIGA family or are already installed or under installation to research reactors of other types. (authors)

Investigation of material removal rate (MRR) and wire wear ratio (WWR) for alloy Ti6Al4 V exposed to heat treatment processing in WEDM and optimization of parameters using Grey relational analysis

International Nuclear Information System (INIS)

Altug, Mehmet

2016-01-01

The study examines the changes of the microstructural, mechanical and conductivity characteristics of the titanium alloy Ti6Al4 V as a result of heat treatment using wire electrical discharge machining, and their effect on machinability. By means of optical microscopy and scanning electron microscopy (SEM), analyses have been performed to determine various characteristics and additionally, microhardness and conductivity measurements have been conducted. Material removal rate (MRR) and wire wear ratio (WWR) values have been determined by using L18 Taguchi test design. The microstructures of the samples have been changed by thermal procedures. Results have been obtained by using the Grey relational analysis (GRA) optimization technique to solve the maximum MRR and minimum WWR values. The best (highest) MRR value is obtained from sample E which was water quenched in dual phase processing. The microstructure of this sample is composed of primary α and α' phases. The best (lowest) WWR value is obtained from sample A.

Investigation of material removal rate (MRR) and wire wear ratio (WWR) for alloy Ti6Al4 V exposed to heat treatment processing in WEDM and optimization of parameters using Grey relational analysis

Energy Technology Data Exchange (ETDEWEB)

Altug, Mehmet [Inonu Univ., Malatya (Turkey). Dept. of Machine and Metal Technologies

2016-11-01

The study examines the changes of the microstructural, mechanical and conductivity characteristics of the titanium alloy Ti6Al4 V as a result of heat treatment using wire electrical discharge machining, and their effect on machinability. By means of optical microscopy and scanning electron microscopy (SEM), analyses have been performed to determine various characteristics and additionally, microhardness and conductivity measurements have been conducted. Material removal rate (MRR) and wire wear ratio (WWR) values have been determined by using L18 Taguchi test design. The microstructures of the samples have been changed by thermal procedures. Results have been obtained by using the Grey relational analysis (GRA) optimization technique to solve the maximum MRR and minimum WWR values. The best (highest) MRR value is obtained from sample E which was water quenched in dual phase processing. The microstructure of this sample is composed of primary α and α' phases. The best (lowest) WWR value is obtained from sample A.

Prague: The City Is the Museum.

Science.gov (United States)

Meilach, Dona Z.

2001-01-01

States that Prague, the capital of the Czech-Republic, is a virtual art museum because of the number of architectural styles and other artworks throughout the city. Explores the various architectural styles that are present in the city from the Gothic monasteries and churches to examples of contemporary styles. (CMK)

Experience of work with radioactive materials and nuclear fuel at the reactor WWR-K

International Nuclear Information System (INIS)

Maltseva, R.M.; Petukhov, V.K.

1998-01-01

In the report there are considered questions concerning the handling with fresh and spent fuel, experimental devices, containing high enriched uranium, being fissile materials of the bulk form, radioisotopes, obtained in the reactor, and radioactive waste, formed during the operation of the reactor, and organization of storage, account and control of radioactive and fissile materials is described. (author)

UrbanTransport Solution An Experience From Prague

African Journals Online (AJOL)

unique firstlady

Based on the result of the research ... associated with road transport like its impact on environment ..... Prague is utilizing a variety of marketing strategies used for many years in ... at strategic metro stations providing customers with maps , time ...

The Allegory of Dynastic Succession on the Facade of the Prague Belvedere (1538-1550)

Czech Academy of Sciences Publication Activity Database

Bažant, Jan

-, č. 2 (2017), s. 269-282 ISSN 0567-8269 Institutional support: RVO:67985955 Keywords : Renaissance art * Ferdinand I * Prague * allegory Subject RIV: AB - History OBOR OECD: History (history of science and technology to be 6.3, history of specific sciences to be under the respective headings)

Research of heat releasing element of an active zone of gaseous nuclear reactor with pumped through nuclear fuel - uranium hexafluoride (UF6)

International Nuclear Information System (INIS)

Batyrbekov, G.; Batyrbekov, E.; Belyakova, E.; Kunakov, S.; Koltyshev, S.

1996-01-01

The purpose of the offered project is learning physics and substantiation of possibility of creation gaseous nuclear reactor with pumped through nuclear fuel-hexafluoride of uranium (Uf6).Main problems of this work are'. Determination of physic-chemical, spectral and optical properties of non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. Research of gas dynamics of laminar, non-mixing two-layer current of gases of hexafluoride of uranium and helium at availability and absence of internal energy release in hexafluoride of uranium with the purpose to determinate a possibility of isolation of hexafluoride of uranium from walls by inert helium. Creation and research of gaseous heat releasing element with pumped through fuel Uf6 in an active zone of research nuclear WWR-K reactor. Objects of a research: Non-equilibrium nuclear - excited plasma of hexafluoride of uranium and its mixtures with other gases. With use of specially created ampoules will come true in-reactor probe and spectral diagnostics of plasma. Calculations of kinetics with the account of main elementary processes proceeding in it, will be carried out. Two-layer non-mixed streams of hexafluoride of uranium and helium at availability and absence of internal energy release. Conditions of obtaining and characteristics of such streams will be investigated. Gaseous heat releasing element with pumped through fuel - Uf6 in an active zone of nuclear WWR-K reactor

Quantitative allochem compositional analysis of Lochkovian-Pragian boundary sections in the Prague Basin (Czech Republic)

Czech Academy of Sciences Publication Activity Database

Weinerová, Hedvika; Hron, K.; Bábek, O.; Šimíček, D.; Hladil, Jindřich

2017-01-01

Roč. 354, JUN 1 (2017), s. 43-59 ISSN 0037-0738 R&D Projects: GA ČR GA14-18183S Institutional support: RVO:67985831 Keywords : compositional analysis * carbonate petrography * multivariate statistics * log-ratio coordinates * Prague Basin * Lower Devonian Subject RIV: DB - Geology ; Mineralogy OBOR OECD: Geology Impact factor: 2.373, year: 2016

State of research and research possibilities of early medieval pottery from the Prague production zone

Czech Academy of Sciences Publication Activity Database

Boháčová, Ivana

2017-01-01

Roč. 55, č. 1 (2017), s. 17-34 ISSN 0066-5924 R&D Projects: GA MK(CZ) DF13P01OVV014 Keywords : early Middle Ages * archaeology * pottery * Prague * chronology * Ceramic Reference Collection Subject RIV: AC - Archeology, Anthropology, Ethnology OBOR OECD: Archaeology

Training reactor deployment. Advanced experimental course on designing new reactor cores

International Nuclear Information System (INIS)

Skoda, Radek

2009-01-01

Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

Participation of Faculty of Mathematics and Physics, Charles University in Prague, in training of personnel for nuclear power

International Nuclear Information System (INIS)

Sterba, F.

1983-01-01

Graduates of application oriented fields of all mathematics and physics specializations of Solid state physics and Nuclear physics work successfully in nuclear power. In the mathematics fields great attention is devoted to optimization, control, process modeling, etc. The subject Solid state physics is subdivided into the following specializations: physics of metals, magnetic properties of the solid state and structural analysis. These specializations educate specialists with a good knowledge of the structure and properties of metal materials. Great attention is devoted to the causes and development of defects, materials creep and the radiation damage of crystal lattices. The nuclear physics specialization Applied nuclear physics deals with the use of nuclear methods in diverse fields and provides basic knowledge in nuclear power generation and the operation of nuclear reactors. The Faculty of Mathematics and Physics of the Charles University in Prague also runs postgraduate study courses in nuclear physics measurement methods, solid state physics, etc. (E.S.)

St George the dragon-slayer at Prague Castle - the eternal pilgrim without a home?

Czech Academy of Sciences Publication Activity Database

Benešovská, Klára

2007-01-01

Roč. 55, č. 1 (2007), s. 28-39 ISSN 0049-5123 Institutional research plan: CEZ:AV0Z80330511 Keywords : Gothic bronze sculpture * St George the dragon -slayer * Martin and George of Cluj * Prague Castle Subject RIV: AL - Art, Architecture, Cultural Heritage http://www.umeni-art.cz/cz/soubory/benesovska.pdf

The oldest Czech fishpond discovered? An interdisciplinary approach to reconstruction of local vegetation in mediaeval Prague suburbs

Czech Academy of Sciences Publication Activity Database

Pokorná, Adéla; Houfková, P.; Novák, J.; Bešta, T.; Kovačiková, L.; Nováková, K.; Zavřel, J.; Starec, P.

2014-01-01

Roč. 730, č. 1 (2014), s. 191-213 ISSN 0018-8158 Grant - others:GA ČR(CZ) GA13-11193S Institutional support: RVO:67985912 Keywords : archaeobotany * archaeozoology * environmental changes * human impact * fishpond * hydrobiology * Prague * the Middle Ages * vegetation diversity Subject RIV: AC - Archeology, Anthropology, Ethnology Impact factor: 2.275, year: 2014

[Physical activity centre VSTJ MEDICINA Prague--rehabilitation for diabetics].

Science.gov (United States)

Fábin, P; Matoulek, M

2007-05-01

Physical activity is the basic non-pharmacological instrument in the treatment of type 2 diabetes. Nevertheless, only a small number of diabetics take regular physical exercise. One of the reasons why diabetics "do not exercise" is that they have little opportunity to try physical stress under expert supervision and to get to know its effects on, for example, sugar levels. It is a very complex matter to define the optimal intensity of physical activity of, for example, a diabetic who suffers from obesity. In 2001 VSTJ MEDICINA Prague opened its first physical activity centre at the First Faculty of Medicine, Charles University in Prague, in cooperation with the Third Internal Clinic and the Institute of Sports Medicine of the First Faculty of Medicine, Charles University in Prague. It now has over 2000 members, of whom around 60% are patients with metabolic syndrome. Over 150 patients exercise every day under the supervision of expert instructors. The main objective of the Physical Activity Centre is to teach patients the correct principles of physical exercise to enable them to continue carrying out their trainers' instructions at home. A correct understanding of the importance of physical exercise and practical experience under the supervision of experienced instructors improves compliance and has a strong effect on the compensation of diabetes, thereby improving the prognoses of these patients.

Local Scaling Properties and Market Turning Points at Prague Stock Exchange

Czech Academy of Sciences Publication Activity Database

Krištoufek, Ladislav

2010-01-01

Roč. 41, č. 6 (2010), s. 1001-1014 ISSN 0587-4254 R&D Projects: GA ČR GA402/09/0965; GA ČR GD402/09/H045 Grant - others:GA UK(CZ) 118310 Institutional research plan: CEZ:AV0Z10750506 Keywords : scaling * Hurst exponent * extreme events Subject RIV: AH - Economics Impact factor: 0.671, year: 2010 http://library.utia.cas.cz/separaty/2010/E/kristoufek-local scaling properties and market turning points at prague stock exchange.pdf

Characterization of anthropogenic influence on the soil cover on selected localities of Prague

Czech Academy of Sciences Publication Activity Database

Žigová, Anna; Šťastný, Martin; Krejčová, J.; Hájek, Pavel

2007-01-01

Roč. 4, č. 3 (2007), s. 39-49 ISSN 1214-9705 R&D Projects: GA AV ČR IAA300130504 Institutional research plan: CEZ:AV0Z30130516; CEZ:AV0Z30460519 Keywords : soil development * clay mineralogy * anthropogenically affected areas of Prague * loess * hot-water extractable carbon Subject RIV: DF - Soil Science http://www.irsm.cas.cz/abstracts/AGG/AGG3_147/5_Zigova.pdf

Nuclear orientation facility at Charles University in Prague

International Nuclear Information System (INIS)

Rotter, M.; Trhlik, M.; Hubalovsky, S.; Srnka, A.; Dupak, J.; Ota, J.; Pari, P.

2000-01-01

A low temperature nuclear orientation facility was installed at Charles University in the laboratory of the Department of Low Temperature Physics on the Faculty of Mathematics and Physics in Prague. The solid state as well as nuclear physics research is pursued on this facility. (author)

Management a marketing sportovní akce: ICF Slalom World Ranking Race Prague 2009

OpenAIRE

Kubričan, Lukáš

2009-01-01

Title: Management and marketing of sport's event: ICF Slalom World Ranking Race Prague 2009 Objectives: Present strengths and weaknesses of ICF Slalom World Ranking Race Prague 2009 based on analyse of recent years and present ideas for its improvement. Methods: Descriptive analysis, SWOT analysis and interview with expert. Results: Conclusion and advices for organizers of sport's events. Key words: Management, marketing, SWOT analysis, descriptive analysis, sport's event, canoe slalom compet...

[Alfred Kohn, professor of histology at German University in Prague].

Science.gov (United States)

Nanka, O; Grim, M

2008-01-01

Prof. Kohn (1867-1959) was the head of the Institute of Histology at the Medical Faculty of German University in Prague for 26 years. In 2007 we commemorated his 140th birthday, and 2009 we will remember the 50th anniversary of his death. He entered the history of medicine by discovery of nature and origin of parathyroid glands and by pioneer research into chromaffin cells and sympathetic paraganglia. Kohn's papers on the pituitary, interstitial cells of testes, and ovaries are also related to endocrinology. All his studies are based on descriptive and comparative histological and embryological observations. Kohn was twice the dean of German Medical Faculty, and a member or honorary member of many important scientific societies. He was repeatedly nominated for Nobel Prize for physiology and medicine. For his Jewish origin he was expelled from Deutsche Gesellschaft der Wissenschaften und Künste für die Tschechoslowakische Republik in 1939 and transported to Terezin ghetto in 1943. After the war he lived in Prague. On the occasion of his 90th birthday he was elected honorary president of Anatomische Gesellschaft and awarded by the Czechoslovak Order of Labour. Alfred Kohn died in 1959. He was one of the outstanding personalities that Prague gave to the world of science.

Catalogue of type specimens of beetles (Coleoptera)deposited in the National Museum, Prague, Czech Republic. Scarabaeoidea: Lucanidae and Passalidae

Czech Academy of Sciences Publication Activity Database

Bezděk, Aleš; Boucher, S.; Hájek, J.

2017-01-01

Roč. 57, č. 1 (2017), s. 279-293 ISSN 0374-1036 Institutional support: RVO:60077344 Keywords : catalogue * types * National Museum Prague Subject RIV: EG - Zoology OBOR OECD: Zoology Impact factor: 0.632, year: 2016

Comparing levels of physical ability and basketball skills of girls in Prague and outsider of Prague

OpenAIRE

Tesaříková, Linda

2017-01-01

The diploma thesis comparing the level of physical abilities and basketball skills of girls in basketball and basketball outfits in Prague and abroad outlines a short history of both world and Czech basketball, the current organization of basketball in the Czech Republic, age specificities of children aged 11, stage of sports training in basketball, Ability. The practical part deals with the question of the level of motor skills of girls at the age of 11, the question of the level of basketba...

Mezinárodní konference ENHR 09 – Prague Changing Housing Markets: Integration and Segmentation

Czech Academy of Sciences Publication Activity Database

Lux, Martin; Vojtková, Michaela

2009-01-01

Roč. 45, č. 5 (2009), s. 1141-1142 ISSN 0038-0288. [ENHR 09 Prague: Changing Housing Markets : Integration and Segmentation. Praha, 28.06.2009-01.07.2009] Institutional research plan: CEZ:AV0Z70280505 Keywords : international conference * housing * integration Subject RIV: AO - Sociology, Demography Impact factor: 0.562, year: 2009 www.enhr2009.cz

Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

Main results and status of the development of LEU fuel for Russian research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

2005-01-01

VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

Tourist Intensity in Capital Cities in Central Europe: Comparative Analysis of Tourism in Prague, Vienna and Budapest

Directory of Open Access Journals (Sweden)

Dumbrovská Veronika

2014-12-01

Full Text Available Urban tourism has become a significant phenomenon of tourism over the last decade. the importance of urban tourism has grown mainly due to the development of transport and information technologies. rapid advancement of low cost airlines and reduction of administrative barriers owing to the expansion of the schengen area caused not only the development of a number of urban destinations, including Prague, but also the growth of new source markets. this paper compares the development of urban tourism in Prague with the situation in Vienna and Budapest in the last decade. the aim of the paper is to describe the main trends of tourism development and the geographic distribution of tourism in Prague in comparison with culturally and historically similar cities - Vienna and Budapest. the analysis shows high load of tourism in Prague and its strong concentration in the old city. this causes congestion in the city centre and an extrusion of residential functions by the functions of tourism. As a result, a tourism ghetto has been formed in the centre of Prague and the urban society has been increasingly dualized.

Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

International Nuclear Information System (INIS)

Hanan, N. A.; Deen, J.R.; Matos, J.E.

2005-01-01

Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

UrbanTransport Solution An Experience From Prague | Jeremiah ...

African Journals Online (AJOL)

This paper examines the urban transport problems in Prague in Czech Republic. Based on the result of the research conducted, it was found that with the collapsed of Communism in Czech Republic, there was an upsurge in the use of private cars which was not possible during communism because the law does not ...

U.S. and foreign breeder reactors

International Nuclear Information System (INIS)

Hill, E.H.

1977-01-01

The running battle between Congress and the Administration over the Clinch River Breeder Reactor Plant (CRBRP) Project has provoked an increased interest in domestic and foreign breeder reactor programs. Perhaps an understanding of the history of breeders here and abroad will serve to place the CRBRP in perspective and allow some analysis of how the U.S. appears on the global canvas. Breeder reactor technology has, for the most part, settled down to concentration on the liquid metal fast breeder reactor (LMFBR). This is the result of 32 years of experience with reactors employing a fast neutron flux and even longer experience with liquid metal coolants. However, a number of U.S. utilities are sponsoring a gas cooled fast reactor program as an alternative technology to the LMFBR. This development program is supported by the U.S. Department of Energy

Study of the cooling systems with S-CO2 for the DEMO fusion power reactor.

Czech Academy of Sciences Publication Activity Database

Veselý, L.; Dostál, V.; Entler, Slavomír

2017-01-01

Roč. 124, November (2017), s. 244-247 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : DEMO * Cooling * Energy conversion * Thermal cycle * Carbon dioxide * SCO2a Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617305719

Test reactor: basic to U.S. breeder reactor development

International Nuclear Information System (INIS)

Miller, B.J.; Harness, A.J.

1975-01-01

Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

Protistology Conferences: The Beginnings. The First International Protozoology Conference (Prague 1961) and the Tribute to Otto Jírovec, its Spiritual Father.

Science.gov (United States)

Vávra, Jiří

2018-01-27

Two events have helped to shape protozoology/protistology as a specific scientific discipline. The first such event was the creation of the Society of Protozoologists in the U.S. in 1947 (and of its Journal of Protozoology, first published in 1954), the second event was the First International Conference on Protozoology, held in 1961 in Prague. The history of the Society of Protozoologists was comprehensively treated by Corliss (1998); the history of the Prague Conference is presented here as reminiscences and personal interpretation of events of the author, who was one of the conference organizers and a member of the organization committee. Special attention is given to the personality and scientific accomplishments of Otto Jírovec, the 1961 conference spiritual father and president. It is concluded that the Prague Conference, while establishing the tradition of protistology meetings, helped protistology to attain its present status as a fundamental science discipline, which discovers and interprets the web of life at one of its, basic, "microbial" levels. Protists literally permeate the earth biosphere and in a way represent the "dark matter" of the living world, still awaiting many discoveries. © 2018 The Author(s) Journal of Eukaryotic Microbiology © 2018 International Society of Protistologists.

Introducing the PET Centre Prague

International Nuclear Information System (INIS)

Belohlavek, O.

2001-01-01

The PET Centre Prague (www.homolka.cz/nm) was established in 1999 as the outcome of a joint project of the public Na Homolce Hospital and the Nuclear Research Institute Rez, plc, the Czech radiopharmaceutical producer. Technical and financial assistance was provided by the International Atomic Energy Agency, which perceived the Centre as its model project that could serve as a guide for the development of PET centres in countries sharing a comparable level of development with the Czech Republic. The article maps the history of the project, its design, workplace lay-out and equipment, radiation protection arrangements and spectrum of the first approx. 3000 investigations. (author)

RUSSIAN FOREIGN HISTORICAL ARCHIVE IN PRAGUE: HISTORY OF FORMATION AND ACTIVITIES

Directory of Open Access Journals (Sweden)

Надежда Александровна Родионова

2015-12-01

Full Text Available The author researches the materials of the Prague Slavonic library to show the struggle of various political groups in exile for the leadership. The analysis of the documents leads to the conclusion that the socialist-revolutionary leadership of the Russian foreign historical archive in Prague was severely criticized by politicians of different lines, seeking to occupy the key positions in the archive to leave "their" history of еmigration to descendants. The examination of the personal records of employees reveals their attitude to the duties and to the archive as a whole. The majority of the staff saw their work in the archive as an honorable mission to preserve the "documented" memory of emigration as a part of the cultural heritage of Russia.

New Plants at Prague Castle and Hradčany in the Early Modern Period. A History of Selected Species

Czech Academy of Sciences Publication Activity Database

Beneš, J.; Čulíková, Věra; Kosňovská, J.; Frolík, Jan; Matiášek, Josef

2012-01-01

Roč. 3, č. 1 (2012), s. 103-114 ISSN 1804-848X Institutional research plan: CEZ:AV0Z80020508 Keywords : Prague Castle * Early Modern Period * archaeobotany Subject RIV: AC - Archeology, Anthropology, Ethnology http://www.iansa.eu/papers/IANSA-2012-01-benes.pdf

The modification of the Rossendorf Research Reactor

International Nuclear Information System (INIS)

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

1990-01-01

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

Gorstian palaeoposition and geotectonic setting of Suchomasty Volcanic Centre (Silurian, Prague Basin, Teplá-Barrandian Unit, Bohemian Massif)

Czech Academy of Sciences Publication Activity Database

Tasáryová, Z.; Schnabl, Petr; Čížková, Kristýna; Pruner, Petr; Janoušek, V.; Rapprich, V.; Štorch, Petr; Manda, Š.; Frýda, J.; Trubač, J.

2014-01-01

Roč. 136, č. 1 (2014), s. 262-265 ISSN 1103-5897 R&D Projects: GA ČR GAP210/10/2351 Institutional support: RVO:67985831 Keywords : basalt geochemistry * Gorstian * palaeolatitude * Prague Basin * Silurian * Suchomasty Volcanic Centre Subject RIV: DE - Earth Magnetism, Geodesy, Geography Impact factor: 1.309, year: 2014

Radon exposure of passengers in the Prague metro

International Nuclear Information System (INIS)

Sabol, J.

1996-01-01

This paper summarises the results of radon concentration monitoring in the carriages and at some stations of the Prague Metro network. The measurements revealed that radon levels in the Metro are relatively low in comparison to those normally encountered in dwellings in the Prague region. On average, the radon concentrations in the air inside the carriages have been found to be about 11 - 12 Bq m -3 while the levels at most stations reached values between 10 and 15 Bq m -3 . The Metro is intensively ventilated by means of powerful blowers and fans; the piston effect of the moving trains also contributes to air exchange. The ventilation rate is typically 3-4 h -1 . The highest rate is in line C, where the air in all underground areas is completely exchanged 6 times within each hour. These results demonstrate that Metro passengers receive about the same effective dose as passengers using surface transport. The doses from radon in the metro are only slightly higher than radon-related doses in the open air, while exposure due to external photon radiation seems to be a few percent lower than dose rates common in typical Czech houses. (author)

Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

2014-01-01

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)

Use of the VR-1 ''Vrabec'' training reactor

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Krops, S.; Polach, S.; Sklenka, L.

1994-01-01

An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

Experiences of activity measurements of primary circuit materials in a WWR-SM research reactor

International Nuclear Information System (INIS)

Elek, A.; Toth, M.; Bakos, L.; Vizdos, G.

1980-01-01

The activity of water and gas samples taken from the primary circuit have been measured nondestructively for more than two years to monitor the technological parameters of the reactor. In the primary water samples 17 fission products and seven activated traces, as well as six radioactive conponents in the gas samples were determined routinely by Ge/Li gamma-spectrometry. (author)

[The Coris, a married couple native to Prague and Nobel laureates].

Science.gov (United States)

Cech, P

2001-01-19

The husband and wife Carl Ferdinand Cori (1896-1984) and Gerty Theresa Radnitz-Cori (1896-1957), two of five Prague-born Nobel laureates (the only ones in medicine), have so much slipped away from the citizens' memory in the course of the half-century totalitarian rule over the country of birth, that hardly anybody knows them nowadays, nothing to say of their relation to Prague. At pains to rescue them from oblivion, a recent search for the lost traces of Coris and their ancestors had revealed a number of hitherto unknown facts that have fundamentally contributed to the Corian genealogy; identification of both forgotten birth-houses (6 Salmovská st., 29 Petrská st.) at long last resulted in placement of memorial tablets (October 26th, 2000) to display the birth-place's pride and gratitude.

Air quality of Prague: traffic as a main pollution source.

Science.gov (United States)

Branis, Martin

2009-09-01

Political and economical transition in the Central and Eastern Europe at the end of eighties significantly influenced all aspects of life as well as technological infrastructure. Collapse of outdated energy demanding industry and adoption of environmental legislation resulted in seeming improvements of urban environmental quality. Hand in hand with modernization the newly adopted regulations also helped to phase out low quality coal frequently used for domestic heating. However, at the same time, the number of vehicles registered in the city increased. The two processes interestingly acted as parallel but antagonistic forces. To interpret the trends in urban air quality of Prague, Czech capital, monthly averages of PM(10), SO(2), NO(2), NO, O(3) and CO concentrations from the national network of automated monitoring stations were analyzed together with long term trends in fuel consumption and number of vehicles registered in Prague within a period of 1992-2005. The results showed that concentrations of SO(2) (a pollutant strongly related to fossil fuel burning) dropped significantly during the period of concern. Similarly NO(X) and PM(10) concentrations decreased significantly in the first half of the nineties (as a result of solid fuel use drop), but remained rather stable or increased after 2000, presumably reflecting rapid increase of traffic density. In conclusion, infrastructural changes in early nineties had a strong positive effect on Prague air quality namely in the first half of the period studied, nevertheless, the current trend in concentrations of automotive exhaust related pollutants (such as PM(10), NO(X)) needs adoption of stricter measures.

U.S. Domestic Reactor Conversion Programs

International Nuclear Information System (INIS)

Woolstenhulme, Eric

2008-01-01

The Conversion Projects Include: the revision of the facilities safety basis documents and supporting analysis, the fabrication of new LEU fuel, the change-out of the reactor core, and the removal of the used HEU fuel (by INL University Fuels Program or DOE-NE). The major entities involved are: the U.S. Nuclear Regulatory Commission, the University reactor department, the fuel and hardware fabricators, the Spent fuel receipt facilities, the Spent fuel shipping services, and the U.S. Department of Energy and their subcontractors. Three major Reactor Conversion Program milestones have been accomplished since 2006: the conversion of the TRIGA reactor at Texas A and M University Nuclear Science Center, the conversion of the University of Florida Training Reactor, and the conversion of the Purdue University Reactor. Four Reactor Conversion Program milestones yet to be accomplished in 2008 and 2009: the Washington State University Nuclear Radiation Center reactor, the Oregon State University TRIGA Reactor, the University of Wisconsin Nuclear Reactor, and the Neutron Radiography Reactor Facility. NNSA is committed to doing things cheaper, better, smarter, safer through a 'Lessons Learned' process. The conversion team assessed each major activity grouping: Project Initiation, Conversion Proposal Development, Fuel Fabrication and Hardware, Core Conversion, and Spent Nuclear Fuel Removal. Issues were identified and recommendations were given

Burnable absorber for the PIK reactor

International Nuclear Information System (INIS)

Gostev, V.V.; Smolskii, S.L.; Tchmshkyan, D.V.; Zakharov, A.S.; Zvezdkin, V.S.; Konoplev, K.A.

1998-01-01

In the reactor PIK design a burnable absorber is not used and the cycle duration is limited by the rods weight. Designed cycle time is two weeks and seams to be not enough for the 100 MW power research reactor equipped by many neutron beams and experimental facilities. Relatively frequent reloading reduces the reactor time on full power and in this way increases the maintenance expenses. In the reactor core fuel elements well mastered by practice are used and its modification was not approved. We try to find the possibilities of installation in the core separate burnable elements to avoid poison of the fuel. It is possible to replace a part of the fuel elements by absorbers, since the fuel elements are relatively small (diameter 5.15mm, uranium 235 content 7.14g) and there are more then 3800 elements in the core. Nevertheless, replacing decreases the fuel burnup and its consumption. In the PIK fuel assembles a little part of the volume is occupied by the dumb elements to create a complete package of the assembles shroud, that is necessary in the hydraulic reasons. In the presented report the assessment of such a replacement is done. As a burnable material Gadolinium was selected. The measurements or the beginning of cycle were performed on the critical facility PIK. The burning calculation was confirmed by measurements on the 18MW reactor WWR-M. The results give the opportunity to twice the cycle duration. The proposed modification of the fuel assembles does not lead to alteration in the other reactor systems, but it touch the burned fuel reprocessing technology. (author)

Lighting up Prague

Energy Technology Data Exchange (ETDEWEB)

Moss, M.

1998-12-31

After almost half a century under the shadow of communism, the Czech Republic is set firm on a path of radical economic reform aimed at restoring its ranking among the most developed states in the world. Through a 145 million pound investment in the independent Czech heat and power producer: Elektrarny Opatovice (EOP), National Power is applying lessons learned during the privatisation of the UK electricity industry to capitalise on the new spirit of liberalisation sweeping through the nation which stands at the very centre of Europe. The National Power Office in Prague is staffed by three men who, together with other specialists from National Power, are in daily contract with their EOP counterparts to share knowledge and best practice in areas ranging from operational maintenance to find management strategies. In a two-way process, EOP is providing National Power with first-hand knowledge of large district heating schemes with efficiencies of up to 90%. EOP operates the coal-fired Opatovice power station supplying 360 MW of electricity and 750 MW of heat. 6 photos.

COMPARISON OF RESEARCH ENGAGEMENT OF PHD STUDENTS AT VARIOUS STUDY PROGRAMS AT CULS PRAGUE: AN INTRODUCTORY STUDY

Directory of Open Access Journals (Sweden)

FLÉGL, Martin

2014-12-01

Full Text Available In an attempt to improve the quality of doctoral studies and the satisfaction of PhD students at the Czech University of Life Sciences Prague (CULS Prague the authors disseminated online questionnaire among all PhD students in May and June 2014. The questionnaire covered areas related to doctoral study, PhD supervisors, doctoral scholarship, research publications, and last but not least, to satisfaction with the doctoral study. In this article responses related to research, such as allocation of time to doctoral studies, allocation of time to research, involvement in research projects and satisfaction with research outputs. The authors provide comparison of all above mentioned domains according to faculties as well as form of doctoral studies at CULS Prague.

Recent U.S. reactor operating experience

International Nuclear Information System (INIS)

Stello, V. Jr.

1977-01-01

A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

Current status of operation and utilization of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Le Van So

2004-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor using the Soviet WWR-SM fuel assembly with 36% enrichment of U-235. It was upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analyses and research purposes. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for physics experiments and training purpose. From the first start-up to the end of December 2002, it totaled about 24,700 hrs of operation and the total energy released was 490 MWd. After 10 years of operation with the core of 89-fuel assembly configuration, in April 1994, the first refueling work was done and the 100-fuel assembly configuration was set-up. The second fuel reloading was executed in March 2002. At present time, the working configuration of the reactor core consists of 104 fuel assemblies. This fuel reloading will ensure efficient exploitation of the reactor for about 3 years with 1200-1300 hrs per year at nominal power. The current status of operation and utilization and some activities related to the reactor core management of the DNRR are presented and discussed in this paper. (author)

The determination of fast neutron fluence in radiation stability tests of steel samples

International Nuclear Information System (INIS)

Hogel, J.; Vespalec, R.

1979-01-01

The activation method is described of determining fast neutron fluence. Samples of steel designed for WWER type reactor pressure vessels were irradiated in the CHOUCA-rigs in the core of the WWR-S reactor. The neutron spectrum was measured by the multiple activation foil method and the effective cross sections of fluence monitors were calculated. The fluences obtained from the reactions 54 Fe(n,p) 54 Mn and 63 Cu(n,α) 60 Co are presented and the method is discussed. (author)

Current status and ageing management of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Nguyen Nhi Dien [Nuclear Research Institute, Dalat (Viet Nam)

2000-10-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor loaded with the Soviet WWR-M2 fuel elements, moderated and cooled by light water. It was reconstructed and upgraded from the former 250 kW TRIGA Mark-II reactor built in 1963. The first criticality of the renovated reactor was in November 1983 and it has been put in regular operation at nominal power since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs every 4 weeks, for radioisotope production, neutron activation analyses and other research purposes. The remaining time is devoted to maintenance work and to short runs for reactor physics studies as well. From its first start-up to the end of 1998, it totaled about 20,000 hrs of operation at nominal power. After ten years of operation, reactor general inspection and refurbishment were implemented in the 1992-1996 period. In April 1994, refueling work was executed with adding of 11 fresh fuel elements to the reactor core. At present, the reactor has been working with 100-fuel element configuration. Corrosion study has been implemented by visual inspection of the reactor pool tank and some other inside components which remain unchanged from the previous TRIGA reactor. The inspections were carried out with the assistance of some experts from other countries. Some visual inspection results have been obtained and the nature of the electrochemical corrosion and related aspects were little by little identified. In this paper, the operation status of the Dalat reactor is presented, and some activities related to the ageing management of the reactor pool tank and its inside components are also discussed. (author)

Current status and ageing management of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien

2000-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW swimming pool type reactor loaded with the Soviet WWR-M2 fuel elements, moderated and cooled by light water. It was reconstructed and upgraded from the former 250 kW TRIGA Mark-II reactor built in 1963. The first criticality of the renovated reactor was in November 1983 and it has been put in regular operation at nominal power since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs every 4 weeks, for radioisotope production, neutron activation analyses and other research purposes. The remaining time is devoted to maintenance work and to short runs for reactor physics studies as well. From its first start-up to the end of 1998, it totaled about 20,000 hrs of operation at nominal power. After ten years of operation, reactor general inspection and refurbishment were implemented in the 1992-1996 period. In April 1994, refueling work was executed with adding of 11 fresh fuel elements to the reactor core. At present, the reactor has been working with 100-fuel element configuration. Corrosion study has been implemented by visual inspection of the reactor pool tank and some other inside components which remain unchanged from the previous TRIGA reactor. The inspections were carried out with the assistance of some experts from other countries. Some visual inspection results have been obtained and the nature of the electrochemical corrosion and related aspects were little by little identified. In this paper, the operation status of the Dalat reactor is presented, and some activities related to the ageing management of the reactor pool tank and its inside components are also discussed. (author)

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

The Sublunary Phaenomena as a Subject of Medieval Academic Discussion. Meteorology and the Prague University Disputationes de Quolibet

Czech Academy of Sciences Publication Activity Database

Kocánová, Barbora

2017-01-01

Roč. 22, č. 1 (2017), s. 72-102 ISSN 1383-7427 Institutional support: RVO:67985955 Keywords : history of meteorology * Middle Ages * Prague University * Aristotle * disputationes de quolibet Subject RIV: AB - History OBOR OECD: History (history of science and technology to be 6.3, history of specific sciences to be under the respective headings) Impact factor: 0.317, year: 2016

Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

Education and research at the VR-1 Vrabec training reactor facility

International Nuclear Information System (INIS)

Matejka, K.

1993-01-01

The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

International Nuclear Information System (INIS)

Matejka, K.; Sklenka, L.

2003-01-01

The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

Photon spectrum behind biological shielding of the LVR-15 research reactor

Energy Technology Data Exchange (ETDEWEB)

Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

2011-07-01

The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

Comment on the in-core measurement in the WWER nuclear power plant

International Nuclear Information System (INIS)

Krett, V.; Dach, K.; Erben, O.

1985-01-01

The activity of the Nuclear Research Institute (NRI) Rez in the field of in-core measurement sensors is described in the paper. The results of comparison and calibration experiments realized on the WWR-S research reactor at the NRI are presented. Measurements with fission calorimeters and SPN detectors carried out in the framework of diagnostic fuel assembly program of WWER NPP reactors are described. Noise measurements with detectors of in-core instrumentation of diagnostic fuel assemblies are also mentioned. Comparison experiments on the WWER-440 NPP reactor are described and the method of function verification of neutron sensors of the in-core control system of these reactors is given. (author)

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

Environmental changes close to the Lower-Middle Devonian boundary; the Basal Choteč Event in the Prague Basin (Czech Republic)

Czech Academy of Sciences Publication Activity Database

Vodrážková, S.; Frýda, J.; Suttner, T. J.; Koptíková, Leona; Tonarová, P.

2013-01-01

Roč. 59, č. 2 (2013), s. 425-449 ISSN 0172-9179 R&D Projects: GA ČR GAP210/10/2351 Institutional support: RVO:67985831 Keywords : Basal Choteč Event * carbon isotope geochemistry * environmental changes * Lower-Middle Devonian * microfacies analysis * Prague Basin Subject RIV: DB - Geology ; Mineralogy Impact factor: 1.338, year: 2013

New results for Palaeozoic volcanic phases in the Prague Basin – magnetic and geochemical studies of Lištice, Czech Republic

Czech Academy of Sciences Publication Activity Database

Elbra, Tiiu; Schnabl, Petr; Tasáryová, Z.; Čížková, Kristýna; Pruner, Petr

2015-01-01

Roč. 64, č. 1 (2015), s. 31-35 ISSN 1736-4728 R&D Projects: GA ČR GAP210/10/2351 Institutional support: RVO:67985831 Keywords : Teplá–Barrandian Unit * Prague Basin * palaeomagnetism * rock magnetism * geochemistry Subject RIV: DE - Earth Magnetism , Geodesy, Geography Impact factor: 0.732, year: 2015

Bioerosion Issue of Ichnos - Collection of Papers from the 4th International Bioerosion Workshop (Prague, August 30-September 3, 2004)

Czech Academy of Sciences Publication Activity Database

Mikuláš, Radek

2006-01-01

Roč. 13, č. 3 (2006), s. 97-97 ISSN 1042-0940. [International Bioerosion Workshop /4./. Prague, 30.08.2004-03.09.2004] Institutional research plan: CEZ:AV0Z30130516 Keywords : bioerosion * conference papers * editorial Subject RIV: DB - Geology ; Mineralogy http://rzblx1.uni-regensburg.de/ezeit/detail.phtml?bibid=CASCR& colors =7&lang=en&jour_id=41560

Vortage passieren lassen: Skizzen zu Transfers zwischen Ort und Zeitbezug in Uwe Johnsons Jahrestagen in Hinsicht auf Prag // Letting Preceding Days Pass: Sketches on the Transfers between Locality and Time-Anchoring in Uwe Johnson’s Anniversaries — with Regard to Prague

Directory of Open Access Journals (Sweden)

Nils Plath

2015-12-01

Full Text Available The paper concentrates on selected passages from Uwe Johnsonʼs opus magnum, Jahrestage (Anniversaries, in which the Prague Spring and the events of 1968 play a key role, and develops some ideas on how narrations of history and critique of language can jointly (as a translation of sorts be understood as a critical approach to what is taken to be the present and the reality of the past. It is argued that Hannah Arendtʼs notion of the unpredictability of the future and of the event variously influenced the modalities of narration as employed by Johnson. The central issue of Jacques Derridaʼs Prague lecture on how a city determines ‘our’ modes of (self-preception and how a location can be grasped and ‘dated’ in writing is put to use in order to develop a singular perspective on some neglected aspects in Johnsonʼs work. Johnson’s novel and its commentaries form a diverse reflection on time, reality, and the media which can affirm its contemporaneity when understood as exemplary in contesting, by the voice of literature, the functionalization of history and its narrative.

Implementation of U.S. Department of Energy physical protection upgrades in Lithuania and Uzbekistan

International Nuclear Information System (INIS)

Haase, M.; Romesberg, L.; Showalter, R.; Soo Hoo, M.S.; Corey, J.; Engling, E.

1996-01-01

Since 1994, the U.S. Department of Energy (DOE) has provided cooperative assistance to the non-nuclear weapons states of the Former Soviet Union. This effort, within DOE's program of Material Protection, Control, and Accounting (MPC ampersand A), identified the Institute of Nuclear Physics (INP) in Uzbekistan and the Ignalina Nuclear Power Plant (INPP) in Lithuania as sites for cooperative MPC ampersand A projects. The INP, located just outside of Tashkent, is the site of a 10-megawatt WWR-SM research reactor. This reactor is expected to remain operational as a major nuclear research and isotope production reactor for Central Asia. The INPP, located 100 kilometers northeast of the capital city of Vilnius, consists of two Russian-made RBMK reactors with a combined power output of 3,000 megawatts (electric). This power plant has been the subject of international safety and security concerns, which prompted DOE's cooperative assistance effort. This paper describes U.S. progress in a multi-national effort directed at implementing physical protection upgrades in Lithuania and Uzbekistan. The upgrades agreed upon between DOE and the INP and between DOE and the INPP have been designed to interface with upgrades being implemented by other donor countries. DOE/INPP upgrade projects include providing training on U.S. approaches to physical protection, access control through the main vehicle portal, a hardened central alarm station, and improved guard force communications. DOE/INP upgrade projects in Uzbekistan include an access control system, a hardened fresh fuel storage vault, an interior intrusion detection and assessment system, and an integrated alarm display and assessment system

Pilot questioning of patological addiction on food - investigation accomplished in selected Prague fast food restaurants and sweetshops

OpenAIRE

Stará, Iveta

2013-01-01

This thesis deals with the phenomenon of food addiction. The research was held among the customers of Prague sweetshops, fast foods and supermarkets. The aim of this work is to find the prevalence of this phenomenon if Prague population, which visits these organizations. The second aim of this work is to deal with prevalence of food dependence in relation with social - demographic readings, such as gender, age, the highest education and a level of month salary. It pays attention to the high i...

Reactor neutron activation analysis on reference materials from intercomparison runs

International Nuclear Information System (INIS)

Pantelica, A.; Salagean, M.

2003-01-01

A review of using the Instrumental Neutron Activation Analysis (INAA) technique in our laboratory to determine major, minor and trace elements in mineral and biological samples from international intercomparison runs organised by IAEA Vienna, IAEA-MEL Monaco, 'pb-anal' Kosice, INCT Warszawa and IPNT Krakow is presented. Neutron irradiation was carried out at WWR-S reactor in Bucharest (short and long irradiation) during 1982-1997 and at TRIGA reactor in Pitesti (long irradiation) during the later period. The following type of materials were analysed: soils, marine sediments, uranium phosphate ore, water sludge, copper flue dust, whey powder, yeast, cereal flour (rye and wheat), marine animal tissue (mussel, garfish and tuna fish), as well as vegetal tissue (seaweed, cabbage, spinach, alfalfa, algae, tea leaves and herbs). The following elements could be, in general, determined: Ag, As, Au, Ba, Br, Ca, Ce, Co, Cr, Cs, Eu, Fe, Hf, Hg, K, La, Lu, Mo, Na, Nd, Ni, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, U, W, Yb and Zn of long-lived radionuclides, as well as Al, Ca, Cl, Cu, Mg, Mn, and Ti of short-lived radionuclides. Data obtained in our laboratory for various matrix samples presented and compared with the intercomparison certified values. The intercomparison exercises offer to the participating laboratories the opportunity to test the accuracy of their analytical methods as well as to acquire valuable Reference Materials/ standards for future analytical applications. (authors)

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

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

Corrosion processes on weathering steel railway bridge in Prague

OpenAIRE

Urban, Viktor; Křivý, Vít; Buchta, Vojtěch

2016-01-01

This contribution deals with experimental corrosion tests carried out on the weathering steel railway bridge in Prague. The basic specific property of the weathering steel is an ability to create in favourable environment a protective patina layer on its surface. Since 1968 weathering steel is used under the name “Atmofix” in the Czech Republic and can be used as a standard structural material without any corrosion protection. The weathering steel Atmofix is mostly used for bridge structures ...

Differences in the spatial patterns of urban tourism in Vienna and Prague

Directory of Open Access Journals (Sweden)

Bálint Kádár

2013-01-01

Full Text Available In Central Europe the two major urban tourism destinations are Vienna and Prague – with both registering the same number of foreign arrivals in 2011. Despite the two cities being similar in their size and range of cultural tourism, they differ significantly in tourists’ spatial distribution and space usage. In Prague, congestion, overcrowding and the mono-functional use of the city centre is well known and documented, whereas in Vienna the city centre hosts a similar number of visitors without conflicts between local functions and tourism. Data obtained from geographically-referenced photography of the two cities uploaded to image-sharing web sites were used to build graphs of the spatial distribution of tourist attractions and routes. Analysing these comparable graphs resulted in some possible explanations regarding the differences in the two cities’ tourist systems. These are mainly related to the morphological layout of the two cities and their divergent approaches to developing urban tourism infrastructures over the past decade.

Travel ticket sales optimization at Prague Main Railway Station

OpenAIRE

Kuběnová, Eva

2010-01-01

This Master's Thesis concerns the travel ticket sales at Prague Main Railway Station. Based on a detailed analysis of provided data, its aim is to put forward suggestions on how to optimize travel ticket sales. Through evaluating the volume of travel tickets sold within the given time period and the number of open cash-desks, along with a cost analysis of sales channels, this thesis reaches its final conclusions. Microsoft Excel was the main analysis tool for compiling graphs. A sound knowled...

Advances in U.S. reactor physics standards

International Nuclear Information System (INIS)

Cokinos, Dimitrios

2008-01-01

The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

Neutron radiography

International Nuclear Information System (INIS)

Hrdlicka, Z.

1977-01-01

Neutron radiography is a radiographic method using a neutron beam of a defined geometry. The neutron source usually consists of a research reactor, a specialized neutron radiography reactor or the 252 Cf radioisotope source. There are two types of the neutron radiography display system, viz., a system producing neutron radiography images by a photographic process or a system allowing a visual display, eg., using a television monitor. The method can be used wherever X-ray radiography is used except applications in the radiography of humans. The neutron radiography unit at UJV uses the WWR-S reactor as the neutron source and both types of the above mentioned display system. (J.P.)

Charged-particle acceleration through laser irradiation of thin foils at Prague Asterix Laser System

International Nuclear Information System (INIS)

Torrisi, Lorenzo; Cutroneo, Maria; Cavallaro, Salvatore; Musumeci, Paolo; Calcagno, Lucia; Wolowski, Jerzy; Rosinski, Marcin; Zaras-Szydlowska, Agnieszka; Ullschmied, Jiri; Krousky, Eduard; Pfeifer, Miroslav; Skala, Jiri; Velyhan, Andreiy

2014-01-01

Thin foils, 0.5–50 μm in thickness, have been irradiated in vacuum at Prague Asterix Laser System in Prague using 10 15–16  W cm −2 laser intensity, 1315 nm wavelength, 300 ps pulse duration and different focal positions. Produced plasmas from metals and polymers films have been monitored in the forward and backward directions. Ion and electron accelerations have been investigated by using Thomson parabola spectrometer, x-ray streak camera, ion collectors and SiC semiconductor detectors, the latter employed in time-of-flight configuration. Ion acceleration up to about 3 MeV per charge state was measured in the forward direction. Ion and electron emissions were detected at different angles as a function of the irradiation conditions. (paper)

Non-proliferation and nuclear disarmament: speech of the president Obama at Prague

International Nuclear Information System (INIS)

Hautecouverture, B.

2009-01-01

Introduced by the Prague speech of april 7 2009, the Obama President program towards the non proliferation and the nuclear disarmament was pointed out by its optimism ambition and determination. But a more detailed lecture shows concurrent positions. The author analyzes the political aspects of the President speech. (A.L.B.)

Time changes of vertical profile of neutron fluence rate in LVR-15 reactor

International Nuclear Information System (INIS)

Viererbl, L.; Stehno, J.; Erben, O.; Lahodova, Z.; Marek, M.

2003-01-01

The LVR-15 reactor is a light water research type reactor, which is situated, in Nuclear Research Institute, Rez near Prague. The reactor is used as a multipurpose facility. For some experiments and material productions, e.g. for homogeneity of silicon resistance in production of radiation doped silicon, the time changes of vertical profile of neutron fluence rate are particularly important. The assembly used for silicon irradiation has two self-powered neutron detectors installed in a vertical irradiation channel in LVR-15 reactor. Vertical profile of thermal neutron fluence rate was automatically scanned during reactor operation. The results of measurements made in 2002 and 2003 with these detectors are presented. A set of vertical profile measurements was made during two 21-days reactor cycles. During the cycle the vertical profile slightly changes both in the position of its maximum and in the shape. The time dependences of the position of profile maximum and the profile width at half maximum during the cycle are given. (author)

[Professor Frantisek Por MD and Professor Robert Klopstock MD, students at Budapest and Prague Faculties of Medicine].

Science.gov (United States)

Mydlík, M; Derzsiová, K

2010-11-01

Professor Frantisek Por MD and Professor Robert Klopstock MD were contemporaries, both born in 1899, one in Zvolen, the other in Dombovar, at the time of Austro-Hungarian Monarchy. Prof. Por attended the Faculty of Medicine in Budapest from 1918 to 1920, and Prof. Klopstock studied at the same place between 1917 and 1919. From 1920 until graduation on 6th February 1926, Prof. Por continued his studies at the German Faculty of Medicine, Charles University in Prague. Prof. Klopstock had to interrupt his studies in Budapest due to pulmonary tuberculosis; he received treatment at Tatranske Matliare where he befriended Franz Kafka. Later, upon Kafka's encouragement, he changed institutions and continued his studies at the German Faculty of Medicine, Charles University in Prague, where he graduated the first great go. It is very likely that, during their studies in Budapest and Prague, both professors met repeatedly, even though their life paths later separated. Following his graduation, Prof. Por practiced as an internist in Prague, later in Slovakia, and from 1945 in Kosice. In 1961, he was awarded the title of university professor of internal medicine at the Faculty of Medicine, Pavol Jozef Safarik University in Kosice, where he practiced until his death in 1980. Prof. Klopstock continued his studies in Kiel and Berlin. After his graduation in 1933, he practiced in Berlin as a surgeon and in 1938 left for USA. In 1962, he was awarded the title of university professor of pulmonary surgery in NewYork, where he died in 1972.

Validation of the Prague C&M classification of Barrett's esophagus in clinical practice

NARCIS (Netherlands)

Alvarez Herrero, Lorenza; Curvers, Wouter L.; van Vilsteren, Frederike G. I.; Wolfsen, Herbert; Ragunath, Krish; Wong Kee Song, Louis-Michel; Mallant-Hent, Rosalie C.; van Oijen, Arnoud; Scholten, Pieter; Schoon, Erik J.; Schenk, Ed B. E.; Weusten, Bas L. A. M.; Bergman, Jacques G. H. M.

2013-01-01

Background and study aims: The Prague C&M classification for Barrett's esophagus has found widespread acceptance but has only been validated by Barrett's experts scoring video sequences. To date, validation has been lacking for its application in routine practice during real-time endoscopy. The aim

Evaluation of neutron flux in the WWR-SM reactor channel and in the irradiating zone of U-150 cyclotron

International Nuclear Information System (INIS)

Sadikov, I.I.; Zinov'ev, V.G.; Sadikova, Z.O.; Salimov, M.I.

2006-01-01

Full text: For effective work of a reactor, and correct planning of experiments related to the reactor irradiation of various materials it is required to control a neutron flux in the given irradiation point for a long irradiation period. For realization of research works on topazes ennobling under irradiation by reactor neutrons as well as by secondary neutrons produced in a cyclotron it is necessary to know the total neutron flux and spectra. To resolve the problem a technique for registration of neutrons with different energy and calculation of a neutrons spectrum in the given irradiation points in reactor channels and in cyclotron behind the nickel target has been developed. Neutron flux density and energy spectra were monitored by use of the following nuclear reactions: 59 Co(n,γ) 60 Co, 197 Au(n,γ) 198 Au, 58 Ni(n,p) 58 Co, 24 Mg(n,p) 24 Na, 48 Ti(n,p) 48 Sc, 46 Ti(n,p) 46 Sc, 54 Fe(n,p) 54 Mn, 89 Y(n,2n) 88 Y, 60 Ni(np) 60 Co. Gamma spectrometer composed of HPGe detector (Rel. Eff. - 15%) and Digital Spectra Analyzer DSA-1000 (Canberra Ind., USA) was used to measure gamma activity of irradiated samples. Acquired gamma spectra were processed by means of Genie 2000 standard software package. The σ(E) functions and neutron spectra were calculated by using the least squares method and approximating the tabular and experimental data with power polynomials. The developed technique was applied for the adjustment of the topazes irradiation regimes in the reactor core and under secondary neutrons flux from a nickel target in the cyclotron. The given technique allows to calculate a logarithmic spectrum of neutrons in a energy range from 0,025 eV up to 12 MeV with the uncertainty of about 10 %. (author)

International Musicological Conference Young Musicology Prague: Czech and European Avant-garde Music of the Early 20th Century, Kabinet hudební historie Etnologického ústavu AV ČR, Praha 5.–8. září 2016

Czech Academy of Sciences Publication Activity Database

Pirner, Jan

2017-01-01

Roč. 54, č. 2 (2017), s. 236-237 ISSN 0018-7003. [International Music ological Conference Young Music ology Prague: Czech and European Avant-garde Music of the Early 20th Century. Prague, 05.10.2016-08.10.2016] Institutional support: RVO:68378076 Keywords : 20th Century * Young Music ology * Conference Subject RIV: AL - Art, Architecture, Cultural Heritage OBOR OECD: Performing arts studies ( Music ology, Theater science, Dramaturgy)

Identification of Chemical Reactor Plant’s Mathematical Model

OpenAIRE

Pyakullya, Boris Ivanovich; Kladiev, Sergey Nikolaevich

2015-01-01

This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

Personal dosimetry in the PET Centre Prague

International Nuclear Information System (INIS)

Janeba, D.; Belohlavek, O.; Hermanska, J.

2001-01-01

This work is focused on radiation protection in the PET Centre Prague. The personal year dose equivalents of physicians, technologists and labtechnologists in the period 1997-2000 are presented. Dose equivalents are listed for each group as collective, mean and maximum dose equivalents and number of people in the evaluated group. There is an increase in the dose equivalents in 1999 when the PET scanner was installed. Later on, when personnel was trained and better local shielding was used, the increase is not much higher even though the number of patients investigated per day doubled. The radiation field measurements showed that the radiation dose equivalent rate outside the controlled area is on the background level of about 0.17-0.18 mSv/hour. (author)

Identification of Chemical Reactor Plant’s Mathematical Model

Directory of Open Access Journals (Sweden)

Pyakillya Boris

2015-01-01

Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

From the history of Czech medieval archaeology. The archaeological excavations of Ivan Borkovský in Loretánské Square in Prague-Hradčany

Czech Academy of Sciences Publication Activity Database

Boháčová, Ivana

2012-01-01

Roč. 57, 1-2 (2012), s. 161-197 ISSN 0003-8180 R&D Projects: GA AV ČR(CZ) IAA800020902 Institutional research plan: CEZ:AV0Z80020508 Institutional support: RVO:67985912 Keywords : Middle Ages * New Ages * cemetery * funeral rite * grave goods * Prague Subject RIV: AC - Archeology, Anthropology, Ethnology

Operating history of U.S. nuclear power reactors

International Nuclear Information System (INIS)

1974-01-01

The operating history of U. S. nuclear power plants through December 31, 1974 has been collected. Included are those nuclear reactor facilities which produce electricity, even if in token amounts, or which are part of a development program concerned with the generation of electricity through the use of a nuclear reactor as a heat source. The information is based on data furnished by facility operators. The charts are plotted in terms of cumulative thermal energy as a function of time. Since only those shutdowns of five days or more are shown, the charts do not give a detailed history of plant operation. They do, however, give an overview of the operating history of a variety of developmental and experimental nuclear power reactors. The data show the yearly gross generation of electricity for each U. S. nuclear plant and, for civilian power plants, information on reactor availability and plant capacity factor. (U.S.)

Research about reactor operator's personability characteristics and performance

Energy Technology Data Exchange (ETDEWEB)

Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

2003-03-01

To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

Prague quadrennial 14.06-20.06/2007. Short play: crisis of forties flood after rain/ Vladimir Anshon

Index Scriptorium Estoniae

Anšon, Vladimir, 1963-

2010-01-01

XI rahvusvahelisest Praha lavastuskunstnike kvadriennaalist (Prague Quadrennial). Eesti rahvuslikust väljapanekust, kuraator Ene-Liis Semper ja Eesti Kunstiakadeemia väljapanekust, kuraator Lilja Blumenfeld

Scutum fidei christianae. The Depiction and Explanation of the Shield of Faith in the Realistic Teaching of Jerome of Prague in the Context of His Interpretation of the Trinity

Czech Academy of Sciences Publication Activity Database

Pavlíček, Ota

2014-01-01

Roč. 62, Suppl. 1 (2014), s. 72-96 ISSN 0015-1831 R&D Projects: GA AV ČR(CZ) KJB900090903 Institutional support: RVO:67985955 Keywords : Jerome of Prague * universals * God * Trinity * scutum fidei * Hussites Subject RIV: AA - Philosophy ; Religion

History of fast reactor development in U.S.A.-I

International Nuclear Information System (INIS)

Ninokata, Hisashi; Sasao, Nobuyki

2007-01-01

History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

Potential Valuation of Route Prague – Trutnov after D11 Completion

Directory of Open Access Journals (Sweden)

Hašek Jakub

2017-11-01

Full Text Available In the Czech Republic, there are many regions with unbalanced competition between bus and rail transport on main transport routes. This is due to differences in the quality of transport infrastructure, the attractiveness of the region, routing traffic routes or economic attractiveness for new investors and industry. The aim of this article is to evaluate the potential for bus operators on the route from Prague to Trutnov after completion of D11 highway project.

Air pollution and respiratory health of children: the PEACE panel study in Prague, Czech Republic.

NARCIS (Netherlands)

Vondra, V.; Branis, M.; Reisova, M.; Maly, M.; Hoek, G.

1998-01-01

A multicentre study (Pollution Effects on Asthmatic Children in Europe (PEACE)) project investigated the relationship between the air pollution and daily variation of respiratory health in children with chronic respiratory symptoms. Data were collected on 66 children in Prague and 68 children in

Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

Science.gov (United States)

Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

2017-11-01

The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

1997-01-01

The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

Kafkův parabolický text Návrat domů. Odraz autorova soukromí a pražských událostí roku 1920 // Kafka’s parabolic text Home-coming. Kafka’s personal life and Prague events of 1920 reflected in a short parabolic story

Directory of Open Access Journals (Sweden)

Jaroslav Březina

2017-10-01

Full Text Available The article focuses on one of Kafka’s short proses of the second half of 1920 — “Home-Coming”. This short parabolic story was created approximately two years after Kafka’s literary pause in the newly established Czechoslovak Republic. The present analysis is based on hypothesis, that Kafka “only” takes up and subverts a traditional mythological theme — in this instance the biblical story of the prodigal son. In this article the text is confronted with contemporary events, the specific situation of the German Jews in Prague and especially episodes of Kafka’s life in order so as to reach for identity the causes and motives of his origin.

Computer simulations for state-of-the-art engineering design of a commercial building in Prague

NARCIS (Netherlands)

Bartak, M.; Drkal, F.; Hensen, J.L.M.; Lain, M.; Schwarzer, J.

2003-01-01

The paper describes the computer simulation work, which was carried out to support the engineering design team of the Luxembourg Plaza building development in Prague. The simulations for this study were based on (1) energy balance models covering the whole building for heating and cooling load

The U.S. Geological Survey's TRIGA® reactor

Science.gov (United States)

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

Prospects of closed-circuit television in detecting surface defects

International Nuclear Information System (INIS)

Kaisler, L. et al.

The use is discussed of closed-circuit television for optical in-service testing of surface defects of nuclear reactors. Experience gained by UJV Rez with in-service testing of the WWR-S reactor is briefly reported. Main attention is devoted to recognizability of defects and to determining the fundamental conditions of the applicability and limitations of the closed-circuit television method. In experiments, resolution of the method was tested and the role of the human factor was assessed in evaluating the results. The need was stressed of thorough training of operators. Based on the experiments conducted, considerations are presented regarding modifications of the individual elements of the tv chain aimed at improved quality of information and a limited role of the observer. (B.S.)

6. Prague colloquium on f-electron systems - PCFES6. Program and abstract booklet

International Nuclear Information System (INIS)

2002-07-01

The 6. Prague Colloquium on f-Electron Systems, held in July 2002, was focused on topics of the current research of electronic structure and material properties of lanthanide and actinide based systems (including magnetic properties and superconductivity); several contributions on transition-metal physics were also included. Altogether 44 lectures and 50 posters were presented. All 94 presentations have been input to INIS. (A.K.)

Spatio-temporal evolution of the 2011 Prague, Oklahoma aftershock sequence revealed using subspace detection and relocation

Science.gov (United States)

McMahon, Nicole D; Aster, Richard C.; Yeck, William; McNamara, Daniel E.; Benz, Harley M.

2017-01-01

The 6 November 2011 Mw 5.7 earthquake near Prague, Oklahoma is the second largest earthquake ever recorded in the state. A Mw 4.8 foreshock and the Mw 5.7 mainshock triggered a prolific aftershock sequence. Utilizing a subspace detection method, we increase by fivefold the number of precisely located events between 4 November and 5 December 2011. We find that while most aftershock energy is released in the crystalline basement, a significant number of the events occur in the overlying Arbuckle Group, indicating that active Meeker-Prague faulting extends into the sedimentary zone of wastewater disposal. Although the number of aftershocks in the Arbuckle Group is large, comprising ~40% of the aftershock catalog, the moment contribution of Arbuckle Group earthquakes is much less than 1% of the total aftershock moment budget. Aftershock locations are sparse in patches that experienced large slip during the mainshock.

Ways of prevention of accidents at atomic reactor

International Nuclear Information System (INIS)

Takibaev, Zh. S.

2000-01-01

The methods proposed to prevent such a move are discussed as well as the scheme of their realization. To improve reactor operation characteristics the safeguard system of quick response is used. Nowadays direct-acting safeguard system (DAS) is to be worked out. It reacts on the main cause of the accident the rapid growth of neutron flux. The time delay of combined gas-liquid DAS unit and fluctuation of nuclear power are calculated. The DAS grid disposed in active zone is developed. Fissile materials are employed because their heating almost immediately follows the growth of neutron flux. There are several systems proposed: uranium bimetal dispersed absorber, uranium hexafluoride liquid absorber (gadolinium solution).Neutronic calculation is done for WWR-1000. The model suggested acts over 0.12 sec. after reactivity swing of 0.003, becomes a 'safety rod' over time delay of 1.49 sec. and cleans itself over 3.0 sec. after.The study presents its improved version. Absorber is injected dose by dose and thus negative reactivity is introduced discretely. Accordingly the same system can act by extracting some parts of fuel from the core. Bimetal safeguard systems are studied. The methods suggested above seem proved in the sense of strengthening nuclear energy development in the future. The problem of DAS and other safeguard systems to prevent reactivity accidents for various reactor types including computer simulation is set to be studied further

Incidence of irradiated foods in the distribution network of Prague

International Nuclear Information System (INIS)

Bohačenko, I.; Kopicova, Z.; Zamecnikova, I.

2005-01-01

The samples, 29 in total, of poultry, rabbit meat, cheese and exotic fruits were taken from the distribution network of Prague. None of the samples was declared as irradiated according to the Decree of the Ministry of Health, CR, No. 133/2004 Sb. The check of their possible exposure to irradiation was made by means of two methods, i.e. the procedure according to EN 1784 (determination of hydrocarbons generated by irradiation using gas chromatography) and the determination of non-bonded o-tyrosine by means of HPLC with electrochemical detection. Neither method brought evidence for the exposure to irradiation, i.e. the purchased foodstuffs concerned were not labelled deceitfully. (author)

Behavior of Photovoltaic System during Solar Eclipse in Prague

Directory of Open Access Journals (Sweden)

Martin Libra

2016-01-01

Full Text Available PV power plants have been recently installed in very large scale. So the effects of the solar eclipse are of big importance especially for grid connected photovoltaic (PV systems. There was a partial solar eclipse in Prague on 20th March 2015. We have evaluated the data from our facility in order to monitor the impact of this natural phenomenon on the behavior of PV system, and these results are presented in the paper. The behavior of PV system corresponds with the theoretical assumption. The power decrease of the PV array corresponds with the relative size of the solar eclipse. I-V characteristics of the PV panel correspond to the theoretical model presented in our previous work.

Accelerator driven reactors and nuclear waste management projects in the Czech Republic

Energy Technology Data Exchange (ETDEWEB)

Janouch, F. [Royal Institute of Technology, Stockholm (Sweden); Mach, R. [Institute of Nuclear Physics, Rez near Prague (Czechoslovakia)

1995-10-01

The Czech Republic is almost the only country in the central Europe which continues with the construction of nuclear power reactors. Its small territory and dense population causes public worries concerning the disposal of the spent nuclear fuel. The Czech nuclear scientists and the power companies and the nuclear industries are therefore looking for alternative solutions. The Los Alamos ATW project had received a positive response in the Czech mass-media and even in the industrial and governmental quarters. The recent scientific symposium {open_quotes}Accelerator driven reactors and nuclear waste management{close_quotes} convened at the Liblice castle near Prague, 27-29. 6. 1994 and sponsored by the Czech Energy Company CEZ, reviewed the competencies and experimental basis in the Czech republic and made the first attempt to formulate the national approach and to establish international collaboration in this area.

Fuel Management at the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

2011-07-01

The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

Complete Genome Sequence of a Burkholderia pseudomallei Strain Isolated from a Pet Green Iguana in Prague, Czech Republic

Science.gov (United States)

Thomas, Prasad; El-Adawy, Hosny; Mertens, Katja; Melzer, Falk; Hnizdo, Jan; Stamm, Ivonne

2017-01-01

ABSTRACT Burkholderia pseudomallei was isolated from pus from an abscess of a pet iguana living in a private household in Prague, Czech Republic. This paper presents the complete genome sequence of B. pseudomallei strain VB976100. PMID:28280033

Network monitoring in the Tier2 site in Prague

International Nuclear Information System (INIS)

Eliáš, Marek; Fiala, Lukáš; Horký, Jirí; Chudoba, Jirí; Kouba, Tomáš; Kundrát, Jan; Švec, Jan

2011-01-01

Network monitoring provides different types of view on the network traffic. It's output enables computing centre staff to make qualified decisions about changes in the organization of computing centre network and to spot possible problems. In this paper we present network monitoring framework used at Tier-2 in Prague in Institute of Physics (FZU). The framework consists of standard software and custom tools. We discuss our system for hardware failures detection using syslog logging and Nagios active checks, bandwidth monitoring of physical links and analysis of NetFlow exports from Cisco routers. We present tool for automatic detection of network layout based on SNMP. This tool also records topology changes into SVN repository. Adapted weathermap4rrd is used to visualize recorded data to get fast overview showing current bandwidth usage of links in network.

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenthal, Murray Wilford [ORNL

2009-08-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

BOOK REVIEW: Tycho Brahe and Prague: Crossroads of European Science

Science.gov (United States)

Sterken, C.; Christianson, J. R.; Hadravová, A.; Hadrava, P.; Solc, M.

2003-01-01

The 16th volume of the Acta Historica Astronomiae is the Proceedings of the International Symposium on the History of Science in the Rudolphine Period. The meeting was held in Prague from 22 to 25 October 2001, on the 400th anniversary of Tycho's sudden death, and was attended by approximately 65 scientists and historians. The volume contains 36 contributions dealing with the life and work of Tycho Brahe, the astronomy of the era, and many cultural aspects of Rudolphine Prague. One of the first papers is an eye-opener on the fact that Tycho Brahe was a cosmologically-driven observer. This is followed by a very illuminating paper on gender roles in science in the late 16th century, with emphasis on the role of Sophie Brahe, Tycho's youngest sister. Several subsequent papers reveal the existence of barely known links between Tycho and his contemporaneous colleagues. These extremely-well documented papers also deal with the broader philosophical investigation he was involved in, viz. meteorology, medicine, astrology, alchemy and even theology. Important names of Tycho's days are Petrus Severinus, Johannes Pratensis, Theophrastus Paracelsus, John Craig, Ursus (Nicolai Reymers Baer) etc. Very illuminating is the information on the relations between Tycho and the Jesuits in Prague, explaining the reason why this order was very supportive of the Tychonic cosmological model. The relationsship with Kepler, and also Kepler's observational activities (after Tycho's death) are highlighted as well as the hideous mode of communication between Galileo and Kepler. More than one paper deals with the accuracy and precision of Tycho's observations, and the causal impact of this accuracy on the scientific revolution. Another study discusses the study of Tycho's handwriting, this paper brings the aditional bonus of a list of accessible works which contain notes by him. One very interesting project was Brahe's proposal to the Republic of Venice to determine the exact latitudes of

Prague-Ruzyne airport visibility data processing results

Science.gov (United States)

Kvicera, Vaclav; Grabner, Martin; Vasicek, Jiri

2010-05-01

Experimental research in the Department of Frequency Engineering in the Czech Metrology Institute (CMI) in Prague, the Czech Republic, is focused on stability of received signal on terrestrial radio and optical communication links. A free space optical (FSO) communication system can provide a high bandwidth solution to last mile broadband access. Dense fog events can cause serious attenuation of optical waves. The availability performances of FSO links are seriously affected by lower atmospheric visibility. Fog seems to be the most important impairment factor for FSO communication links. Therefore, our experimental research is also focused on processing meteorological data to obtain basic information for the assessment of availability performances of FSO links. The visibility measured at airports provides a good estimate for the assessment of fog impairment. The visibility data that is measured and stored in meteorological stations or airports is usually used for the calculation of specific attenuation due to fog in accordance with either the relevant ITU-R Recommendation or other common methods. Fog visibility data obtained by the Vaisala transmissometer at the Prague-Ruzyne airport from January 1996 to December 2008 (13 years of observation) was statistically processed over the individual years, the individual months and the individual hours and the following cumulative distributions (CDs) of visibility were obtained: 1) CDs of visibility for individual years, 2) CDs of visibility for individual months over the whole period of processing, 3) CDs of visibility for the worst hour over the whole period of processing, and 4) CDs of visibility for the average year and the average worst month over the entire 13 year period of processing. It can be concluded that the frequency of fog events shows a great year-to-year variability in accordance with the frequency of individual synoptic situations occurring during the year. Fog occurred most frequently in November and

Assessment methodology applicable to safe decommissioning of Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Baniu, O.; Vladescu, G.; Vidican, D.; Penescu, M.

2002-01-01

The paper contains the results of research activity performed by CITON specialists regarding the assessment methodology intended to be applied to safe decommissioning of the research reactors, developed taking into account specific conditions of the Romanian VVR-S Research Reactor. The Romanian VVR-S Research Reactor is an old reactor (1957) and its Decommissioning Plan is under study. The main topics of paper are as follows: Safety approach of nuclear facilities decommissioning. Applicable safety principles; Main steps of the proposed assessment methodology; Generic content of Decommissioning Plan. Main decommissioning activities. Discussion about the proposed Decommissioning Plan for Romanian Research Reactor; Safety risks which may occur during decommissioning activities. Normal decommissioning operations. Fault conditions. Internal and external hazards; Typical development of a scenario. Features, Events and Processes List. Exposure pathways. Calculation methodology. (author)

Radiation research of materials using irradiation capsules

International Nuclear Information System (INIS)

Chamrad, B.

1976-01-01

The methods are briefly characterized of radiation experiments on the WWR-S research reactor. The irradiation capsule installed in the reactor including the electronic instrumentation is described. Irradiated samples temperature is stabilized by an auxiliary heat source placed in the irradiation space. The electronic control equipment of the system is automated. In irradiation experiments, experimental and operating conditions are recorded by a digital measuring centre with electric typewriter and paper tape data recording and by an analog compensating recorder. The irradiation experiment control system controls irradiated sample temperature, the supply current size and the heating element temperature of the auxiliary stabilizing source, inert and technological pressures of the capsule atmosphere and the thermostat temperature of the thermocouple junctions. (O.K.)

Inelastic scattering of 275 keV neutrons by silver

International Nuclear Information System (INIS)

Litvinsky, L.L.; Zhigalov, Ya.A.; Krivenko, V.G.; Purtov, O.A.; Sabbagh, S.

1997-01-01

Neutron total, elastic and inelastic scattering cross-scattering of Ag at the E n = 275 KeV neutron energy were measured by using the filtered neutron beam of the WWR-M reactor in Kiev. The d-neutron strength function S n2 of Ag was determined from the analysis of all available data in the E n ≤ keV energy region on neutron inelastic scattering cross-sections with excitation of the first isomeric levels I π m = 7/2 + , E m ∼ 90 keV of 107,109 Ag: S n2 = (1.03 ± 0.19) · 10 -4 . (author). 10 refs, 3 figs

Komunikační strategie Mercedes-Benz Prague Fashion Weekendu v letech 2013-2014

OpenAIRE

Pavlík, Mikuláš

2015-01-01

The bachelor thesis "The Communication Strategy of Mercedes-Benz Prague Fashion Weekend in 2013-2014" analyses the overall functioning and transformation of the communication strategy of this key social and business event that promotes the Czech fashion industry. The observed period of two years focuses on the change of leadership and the entrance of the key sponsor and also the titular partner - Mercedes-Benz, which as part of its strategy supports fashion weeks on an international level. In...

The laboratory of quantitative methods in historic monument research at the CTU Prague

International Nuclear Information System (INIS)

Musilek, L.; Cechak, T.; Kubelik, M.; Pavelka, K.; Pavlik, M.

2001-01-01

A 'Laboratory of Quantitative Methods in Historic Monument Research' has been established at the Department of Dosimetry and Application of Ionizing Radiation of the CTU Prague. Its primary orientation is the investigation of historic architecture, although other objects of art can also be, investigated. In the first phase, one investigative method was established for each of the above groups: X-ray fluorescence as the analytic method, thermoluminescence for dating and photogrammetry for surveying. The first results demonstrate the need and usefulness of these methods for investigations in the rich architectural heritage of the Czech Republic.

Enhancement the physical protection system of the WWR-SM reactor at Institute of Nuclear Physics of Academy of Science of the Republic of Uzbekistan

International Nuclear Information System (INIS)

Karabaev, Kh.Kh.; Rakhimbaev, A.T.; Rakhmanov, A.B.; Salikhbaev, U.S.; Yuldashev, B.S.

2004-01-01

Full text: Joining of the Republic of Uzbekistan to Non-Proliferation Treaty required the revision of nuclear fuel protection system and radioactive sources from illegal access in all stages of work with nuclear materials. One of the primary technical actions of ensuring non-proliferation of nuclear materials is physical protection. The project was worked out on upgrading and enhancement of the physical protection of the reactor building. In cooperation with Sandia National Laboratory and support of the Department of Energy (DOE) USA The first stage of the physical protection upgrading provided for fresh fuel protection: - the new fresh fuel storage room was built and equipped with the modern control and detection system, - the reactor building was equipped with detection devices and access control, - the central alarm station (CAS) has been built and equipped with computer control and observing system, - code access system has been implemented. The first stage of upgrading of physical protection system was accomplished for 4 months, and put into operation in 1996. The second stage of physical protection system modernization included the construction of the second barrier of the physical protection, equipping it with observation and control devices and also extension of the CAS. The perimeter around the reactor building was cleaned from trees, bushed and in a short time a two-fence barrier was erected. The access control point provided the secured intensified control of the access to the reactor territory. The physical protection system was supplied with equipment for safeguard and TV observation of perimeter, access control to the territory of the reactor: - the CAS was extended and computer observation control system was upgraded, - the badge station has been constructed, equipped and set up, - all doors, windows, reactor hall gate have been replaced by strengthened metal ones, - uninterruptible power supply (UPS) and diesel-generator have been installed, - the

Final report. U.S. Department of Energy University Reactor Sharing Program

Energy Technology Data Exchange (ETDEWEB)

Bernard, John A

2003-01-21

Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

Calibration of a special neutron dosemeter based on solid-state track detectors and fission radiators in various neutron fields

International Nuclear Information System (INIS)

Doerschel, B.; Krusche, M.; Schuricht, V.

1980-01-01

The calibration of a personnel neutron dosemeter in different neutron fields is described. The badge-like dosemeter contains 5 detectors: polycarbonate foil (10 μm, Makrofol KG), 232 Th, natural uranium, natural uranium with boron, and natural uranium with cadmium. Detector sensitivity and calibration factors have been calculated and measured in radiation fields of 252 Cf fission neutrons, WWR-S reactor neutrons with and without Cd and Fe shielding, 3-MeV (d,t) generator neutrons, and 238 PuBe neutrons. Measurement range and achievable accuracy are discussed from the point of view of applying the dosemeter in routine and emergency uses

U.S. uranium supply to the research and test reactor community

International Nuclear Information System (INIS)

Parker, Elaine M.

2002-01-01

From the 1950s through the early 1990s, the U.S. Department of Energy (DOE) was the primary supplier of low enriched uranium (LEU) and highly enriched uranium (HEU) to research and test reactors worldwide. The formerly called Y-12 Plant in Oak Ridge, Tennessee, was put into operational stand down in 1994 due to inadequate safety documentation. This paper will discuss the re-start of the Y-12 Plant and its current capabilities. Additionally, the paper will address recent changes within the DOE, with the creation of the National Nuclear Security Administration (NNSA). It will show how the change to NNSA and an organizational re-alignment has improved efficiencies. NNSA is committed to operate its sales program so that it is complementary to, and in support of, the Reduced Enrichment for Research and Test Reactors (RERTR) and Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Return Programs. The NNSA is committed to provide an assurance of competitively-priced, high-quality uranium supply to the research and test reactor community under long-term contracts. This paper will discuss some of NNSA's recent successes in long-term contracting and meeting deliveries. (author)

Safe decommissioning of the Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Garlea, C.; Garlea, I.; Kelerman, C.; Rodna, A.

2002-01-01

The VVR-S Romania research reactor was operated between 1957-1997, at 2 MW nominal power, for research and radioisotopical production. The detailed decommissioning plan was developed between 1995-1998, in the frame of the International Atomic Energy Agency Technical assistance project ROM/9/017. The proposed strategy agreed by the counterpart as well as international experts was stage 1. In 1997, an independent analysis performed by European Commission experts, in the frame of PHARE project PH04.1/1994 was dedicated to the 'Study of Soviet Design Research Reactors', had consolidated the development of the project emphasizing technical options of safe management for radioactive wastes and VVR-S spent fuel. The paper presents the main technical aspects as well as those of social impact, which lead to the establishment of strategy for safe management of decommissioning. Technical analysis of the VVR-S reactor and associated radwaste facilities (Radioactive Waste Treatment Plant - Magurele and National Repository Baita-Bihor) proved the possibility of the classical method utilization for dismantling of the facility and treatment-conditioning-disposal of the arrised wastes in safe conditions. The decommissioning plan at stage 2 has been developed based on radiological safety assessment, evaluation of radwaste inventory (removed as well as preserved on site), cost analysis and environmental impact. Technical data were provided by the R and D programme including neutron calculations and experiments, radiological characterizing (for facility and its influence area), seismic analysis and environmental balance during the operation and after shut down of the reactor. A special chapter is dedicated to regulatory issues concerning the development of decommissioning under nuclear safety. Based on the Fundamental Norms of Radiological Safety, the Regulatory Body defined the clearance levels and safety criteria for the process. The development of National Norms for the

Prague Overview Week, September 13th-20th, 2003

CERN Multimedia

Smith, K

The week already began on Saturday for a number of system and sub-system meetings, but got into full swing on the Sunday, with a marathon session on how to tackle the challenging task of commissioning the full ATLAS spectrometer. The seeds of lots of ideas have already germinated and should begin to flower soon. Simulation studies of cosmic ray fluxes (and even early experimental measurements in Pit 1) have shown useful rates for the pre-beam phase of commissioning, and single beam measurements will also offer useful opportunities, in particular for trigger timing. Commissioning is clearly a topic which will continue to grow in importance during the next few years. On Monday we were welcomed by our Czech hosts to the first "official" session of the Overview Week, held in the Department of Mechanical Engineering of the Czech Technical University, (CTU), the oldest Technical University in Europe. ATLAS (Prague) includes groups from the CTU, the even longer established Charles University and the Czech Acad...

Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

International Nuclear Information System (INIS)

Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

2016-01-01

Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

[Cytostatic hyperthermic isolated limb perfusion (HILP) in VFN (General Faculty Hospital in Prague)].

Science.gov (United States)

Spacek, M; Mitás, P; Lacina, L; Krajsová, I; Hodková, G; Salmay, M; Spunda, R; Brlicová, L; Lindner, J

2011-01-01

Hyperthermic isolated limb perfusion (HILP) is a standardized method of treatment in selected patients with in-transient locoregional recurrence/methastasis of melanoma or, some other soft tissue tumors (incl. sarcoma etc.) Authors present history and current status of this treatment modality in General University Hospital in Prague. During one year period (7/2009-6/2010) 10 patients were indicated for this procedure. We performed 13 procedures (3x redo), 11 in lower extremity and 2 in upper extremity. There was no serious complication in this cohort of patiens. Multidisciplinar approach is indicated in melanoma patients care.

Facilities available for actinide research in Prague

International Nuclear Information System (INIS)

Sechovský, V.

2014-01-01

Since June 2012 the Prague group at the Charles University operates a Czech research infrastructure Magnetism and Low Temperature Laboratories (MLTL - http://mltl.eu orhttp://lmnt.cz)which is financially supported by the Government of Czech Republic. The main mission of MLTL is to provide broad scientific community unique possibilities for comprehensive experimental studies of physical phenomena and properties of materials in multiextreme conditions.MLTL offer open access to a wide range of experimental facilities for sample preparation (SSE refinement of staring metals, synthesis of bulk polycrystals, growth of single crystals), characterization (XRD, SEM + EDX) and measurements of various physical properties in high magnetic fields up to 20 T, temperatures from 30 mK to 1000Kand external pressures up to 25 GPa). Anybody can apply for experimental time with his proposal on the user portal of http://mltl.eu. The main strategic objective is the excellence of the infrastructure on the international scale. Therefore the MLTL Panel evaluation the proposals and allocation of experimental time is based primarily on the quality of intended research. The proposals of students for experiments needed for their theses are promoted within the evaluation process. The research opportunities offered by MLTL will be demonstrated during the lecture with emphasis on methodology

Practical experience in the application of quality control in water-reactor fuel fabrication

International Nuclear Information System (INIS)

Vollath, D.

1984-07-01

Highly industrialized countries have gained vast experience in manufacturing water reactor fuel. Manufacturing is followed by a stringent system of quality assurance and quality control. The Seminar on Practical Experience in the Application of Quality Control in Water-Reactor Fuel Fabrication provided a forum for an exchange of information on methods and systems of quality assurance and quality control for reactor fuel. In addition, many developing countries which have started or intend to set up a nuclear fuel industry are interested in the application of quality assurance and quality control. This meeting has been preceded by two different series of conferences: the IAEA meetings 1976 in Oslo, 1978 in Prague and 1979 in Buenos Aires, and the Karlsruhe meetings on Characterization and Quality Control of Nuclear Fuel held in 1978 and 1981. Quality control and quality assurance has many different facets. Unlike the purely technical aspects, covered by the Karlsruhe conference series, the IAEA meetings always relate to a wider field of topics. They include governmental regulations and codes for practical quality assurance. This volume contains the papers presented at the seminar and a record of the discussions. (orig.)

Precise position of the Basal Choteč event and evolution of sedimentary environments near the Lower–Middle Devonian boundary: The magnetic susceptibility, gamma-ray spectrometric, lithological, and geochemical record of the Prague Synform (Czech Republic)

Czech Academy of Sciences Publication Activity Database

Koptíková, Leona

2011-01-01

Roč. 304, 1/2 (2011), s. 96-112 ISSN 0031-0182 R&D Projects: GA AV ČR KJB307020602; GA AV ČR IAAX00130702 Institutional research plan: CEZ:AV0Z30130516 Keywords : magnetic susceptibility * Gamma-ray spectrometry * Prague Synform * Lower–Middle Devonian Subject RIV: DB - Geology ; Mineralogy Impact factor: 2.392, year: 2011

Performance of Continuous Micro Photo Reactor – Comparison with Batch Process

Czech Academy of Sciences Publication Activity Database

Drhová, Magdalena; Hejda, S.; Křišťál, Jiří; Klusoň, Petr

2012-01-01

Roč. 42, SI (2012), s. 1365-1372 E-ISSN 1877-7058. [International Congress of Chemical and Process Engineering CHISA 2012 and 15th Conference PRES 2012 /20./. Prague, 25.08.2012-29.08.2012] Institutional support: RVO:67985858 Keywords : continuous microreactor * photooxidation * phtalocyanine Subject RIV: CI - Industrial Chemistry, Chemical Engineering

Personal exposures to NO2 in the EXPOLIS-study: relation to residential indoor, outdoor and workplace concentrations in Basel, Helsinki and Prague

International Nuclear Information System (INIS)

Kousa, A.; Rotko, T.; Alm, S.; Monn, C.

2001-01-01

Personal exposures, residential indoor, outdoor and workplace levels of nitrogen dioxide (NO 2 ) were measured for 262 urban adult (25-55 years) participants in three EXPOLIS centres (Basel, Switzerland; Helsinki, Finland; and Prague, Czech Republic) using passive samplers for 48-h sampling periods during 1996-1997. The average residential outdoor and indoor NO 2 levels were lowest in Helsinki (24 ± 12 and 18 ± 11 μgm -3 , respectively), highest in Prague (61 ± 20 and 43 ± 23μgm -3 ), with Basel in between (36 ± 13 and 27± 13μgm -3 ). Average workplace NO 2 levels, however, were highest in Basel (36 ± 24μgm -3 ), lowest in Helsinki (27 ± 15μgm -3 ), with Prague in between (30 ± 18μgm -3 ). A time-weighted microenvironmental exposure model explained 74% of the personal exposure variation in all centre and in average 88% of the exposures. Log-linear regression models, using residential outdoor measurements (fixed site monitoring) combined with residential and work characteristics (i.e. work location, using gas appliances and keeping windows open), explained 48% (37%) of the personal NO 2 exposure variation. Regression models based on ambient fixed site concentrations alone explained only 11-19% of personal NO 2 exposure variation. Thus, ambient fixed site monitoring alone was a poor predictor for personal NO 2 exposure variation, but adding personal questionnaire information can significantly improve the predicting power. (Author)

The Ring Monstrance from the Loreto treasury in Prague: handheld Raman spectrometer for identification of gemstones.

Science.gov (United States)

Jehlička, Jan; Culka, Adam; Baštová, Markéta; Bašta, Petr; Kuntoš, Jaroslav

2016-12-13

A miniature lightweight portable Raman spectrometer and a palm-sized device allow for fast and unambiguous detection of common gemstones mounted in complex jewels. Here, complex religious artefacts and the Ring Monstrance from the Loreto treasury (Prague, Czech Republic; eighteenth century) were investigated. These discriminations are based on the very good correspondence of the wavenumbers of the strongest Raman bands of the minerals. Very short laser illumination times and efficient collection of scattered light were sufficient to obtain strong diagnostic Raman signals. The following minerals were documented: quartz and its varieties, beryl varieties (emerald), corundum varieties (sapphire), garnets (almandine, grossular), diamond as well as aragonite in pearls. Miniature Raman spectrometers can be recommended for common gemmological work as well as for mineralogical investigations of jewels and cultural heritage objects whenever the antiquities cannot be transported to a laboratory.This article is part of the themed issue 'Raman spectroscopy in art and archaeology'. © 2016 The Author(s).

Land Registers in the Czech Lands and their crucial role in studying the history of the landscape Cadastral Maps as the sources of the Historic Towns Atlas of the Czech Republic and their place in the on-line GIS Portal of Prague Historical Cartography and Iconography

Czech Academy of Sciences Publication Activity Database

Chodějovská, Eva

2012-01-01

Roč. 31, č. 4 (2012), s. 277-288 ISSN 2035-8733. [I catasti e la storia dei luoghi. Cagliari, 12.10.2012-13.10.2012] Institutional support: RVO:67985963 Keywords : Old maps * Web Map Portal * GIS * Prague

Evaporation from water surfaces in urban environments, using Prague and Pilsen (Czech Republic as examples

Directory of Open Access Journals (Sweden)

Knozová Gražyna

2016-12-01

Full Text Available The subject of this study is an evaluation of the amount of evaporation from water surfaces (VVH, measured using EWM devices in two cities of different sizes, and located approximately 80 km from each other – Prague and Pilsen. The results were analyzed in the context of urban phenomena, which are pronounced especially in Prague, and also in the context of meteorological and morphological conditions in those locations. It was found that higher amounts of evaporation were measured at the meteorological station in Pilsen. The difference between the average sum of VVH per season (1st May to 30th September between 2005 and 2014 for the two locations is 33.3 mm. The difference between daily average values was 0.2 mm. Given the suburban nature of the two locations where measurements were taken, it was not possible to draw any conclusions about the effect of the urban heat island on the rate of evaporation and values of VVH. Factors significantly influencing VVH are surface roughness, which is higher in urban environments than in open landscapes. Based on the results it was concluded that at both a regional and a local scale, the rate of evaporation is more affected by wind speed than thermal conditions. The measured VVH values differ, not just because of the urban dimension of the two cities compared, but especially as a result of different topoclimatic location of the two stations.

LPHYS'13: 22nd International Laser Physics Workshop (Prague, 15-19 July 2013)

Science.gov (United States)

Yevseyev, Alexander V.

2013-04-01

The 22nd annual International Laser Physics Workshop (LPHYS'13) will be held from 15-19 July 2013 in the city of Prague, Czech Republic, at the Hotel Krystal and Czech Technical University hosted this year by the Institute of Physics ASCR and Czech Technical University in Prague. LPHYS'13 continues a series of workshops that took place in Dubna, 1992; Dubna/Volga river tour, 1993; New York, 1994; Moscow/Volga river tour (jointly with NATO SILAP Workshop), 1995; Moscow, 1996; Prague, 1997; Berlin, 1998; Budapest, 1999; Bordeaux, 2000; Moscow, 2001; Bratislava, 2002; Hamburg, 2003; Trieste, 2004; Kyoto, 2005; Lausanne, 2006; León, 2007; Trondheim, 2008; Barcelona, 2009; Foz do Iguaçu, 2010; Sarajevo, 2011; and Calgary, 2012. The total number of participants this year is expected to be about 400. In the past, annual participation was typically from over 30 countries. 2013 Chairmen: Miroslav Jelinek (Czech Republic) and Pavel P Pashinin (Russia) LPHYS'13 will offer eight scientific section seminars and one general symposium: Seminar 1 Modern Trends in Laser Physics Seminar 2 Strong Field & Attosecond Physics Seminar 3 Biophotonics Seminar 4 Physics of Lasers Seminar 5 Nonlinear Optics & Spectroscopy Seminar 6 Physics of Cold Trapped Atoms Seminar 7 Quantum Information Science Seminar 8 Fiber Optics Symposium Extreme Light Technologies, Science and Applications Abstract of your presentation A one-page abstract should contain: title; list of all co-authors (the name of the speaker underlined); affiliations; correspondence addresses including phone numbers, fax numbers, e-mail addresses; and the text of the abstract. Abstracts should be sent to the following co-chairs of the scientific seminars and the symposium: Kirill A Prokhorov (Seminar 1) E-mail: cyrpro@gpi.ru Mikhail V Fedorov (Seminar 2) E-mail: fedorov@ran.gpi.ru Sergey A Gonchukov (Seminar 3) E-mail: gonchukov@mephi.ru Ivan A Shcherbakov (Seminar 4) E-mail: gbufetova@lsk.gpi.ru Vladimir A Makarov (Seminar 5) E

U. S. Utility Leadership in Requirements For Passive Reactors

International Nuclear Information System (INIS)

Kim, Jcng H.; Layman, William H.

1991-01-01

Utility leadership from both U.S. utilities and international utilities, is a key element in the U. S. Advanced Light Water Reactor Program. International utilities have played a very import Design reviews by the utilities participating in the ALRR Program will ensure that all of the utility requirements are met while design work is being carried out. Our mission is to achieve NRC certification of designs that reflect the needs of the utilities and we believe that this will play an important role in the resurgence of nuclear plant construction in the United States. As stated in the Nuclear Power Oversight Committee's Strategic Plan For Building New Nuclear Power Plants : 'The extensive operating experience with today's light water reactors (LWRs), and the promise shown in recent technical developments, leads the industry to the conclusion that the next nuclear plants ordered in the United States will be advanced light water reactors (A LWRs). Two types are under development : units of large output (1300 MW) called 'evolutionary' A LWRs and units of mid-size output (600 MW) called 'Passive' A LWRs. The term 'passive' refers to the safety features which depend more on natural processes such as gravity and buoyancy than on powered equipment such as pumps

Effective dose to patient during cardiac interventional procedures (Prague workplaces)

International Nuclear Information System (INIS)

Stisova, V.

2004-01-01

The aim of this study was to assess effective dose to a patient during cardiac procedures, such as coronary angiography (CA) and percutaneous transluminal angioplasty (PTCA). Measurements were performed on 185 patients in four catheterisation laboratories in three hospitals in Prague using the dose area product (DAP) meter. Calculations of surface and effective dose were performed with Monte-Carlo-based program PCXMC. The mean DAP value per procedure determined in all workplaces ranged between 25.0 and 54.5 Gy cm 2 for CA and 43.0-104.5 Gy cm 2 for PTCA. In three cases, the surface dose exceeded the 2 Gy level for occurrence of transient erythema. The mean effective dose per procedure in an workplaces was determined to be in the range of 2.7-8.8 mSv for CA and 5.7-15.3 mSv for CA + PTCA combined. The results presented are comparable with those published by other authors. (authors)

Current status of operation and utilization of the Dalat Research Reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi

2006-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool-type reactor using the HEU (36% enrichment) WWR-M2 fuel assemblies. It was renovated and upgraded from the USA 250 kW TRIGA Mark-II reactor. The first criticality of the renovated reactor was in the 1st November 1983 and its regular operation at nominal power of 500 kW has been since March 1984. The DNRR is operated mainly in continuous runs of 100 hrs, once every 4 weeks, for radioisotope production, neutron activation analysis, scientific research and training. The remaining time between two continuous runs is devoted to maintenance activities and also to short run for reactor physics and thermal hydraulics experiments. From the first start-up to the end of December 2004, it totaled about 27,253 hrs of operation and the total energy released was about 543 MWd. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 fuel assemblies (FA). The 11 new FAs were added in the core periphery, at previous beryllium element locations. After reloading the working configuration of reactor core consisted of 100 FAs. The second fuel reloading was executed in March 2002. The 4 new FAs were added in the core periphery, at previous beryllium element locations. The working configuration of 104 FAs ensured efficient exploitation of the DNRR at nominal power for about 3000 hrs since March 2002. In order to provide excess reactivity for the reactor operation without the need to discharge high burned FAs, in June 2004, the fuel shuffling of the reactor core was done. 16 FAs with low burn-up from the core periphery were moved toward the core center and 16 FAs with high-burn-up from the core center were moved toward the core periphery. This operation provided additional reactivity of about 0.85 β eff that the current reactor configuration using re-shuffled HEU fuel is expected to allow normal operation until June 2006. In 1999, the request of returning to Russia HEU fuels from foreign

Full instantaneous traversal rupture of the primary loop pipeline

International Nuclear Information System (INIS)

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

2010-01-01

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

A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

Energy Technology Data Exchange (ETDEWEB)

Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

2013-06-15

HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

Computer control system synthesis for nuclear power plants through simplification and partitioning of the complex system model into a set of simple subsystems

International Nuclear Information System (INIS)

Zobor, E.

1978-12-01

The approach chosen is based on the hierarchical control systems theory, however, the fundamentals of other approaches such as the systems simplification and systems partitioning are briefly summarized for introducing the problems associated with the control of large scale systems. The concept of a hierarchical control system acting in broad variety of operating conditions is developed and some practical extensions to the hierarchical control system approach e.g. subsystems measured and controlled with different rates, control of the partial state vector, coordination for autoregressive models etc. are given. Throughout the work the WWR-SM research reactor of the Institute has been taken as a guiding example and simple methods for the identification of the model parameters from a reactor start-up are discussed. Using the PROHYS digital simulation program elaborated in the course of the present research, detailed simulation studies were carried out for investigating the performance of a control system based on the concept and algorithms developed. In order to give a real application evidence, a short description is finally given about the closed-loop computer control system installed - in the framework of a project supported by the Hungarian State Office for Technical Development - at the WWR-SM research reactor where the results obtained in the present IAEA Research Contract were successfully applied and furnished the expected high performance

Nuclear reactor safety program in U.S. Department of Energy and future perspectives

International Nuclear Information System (INIS)

Song, Y.T.

1987-01-01

The U.S. Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and impliment DOE orders. Teams of experts in the Depatment, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The application of artificial intelligence (AI) technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs. (author)

Final report on the IAEA research contracts No. 1194/RB, 1194/R1/RB and 1194/R2/RB

International Nuclear Information System (INIS)

Zobor, E.; Janosy, J.S.; Szentgali, A.

1980-09-01

The final report summarizes the research activities made in the framework of the IAEA Research Contracts No. 1194/RB, 1194/R1/RB and 1194/R2/RB. A multilevel hierarchical control system is treated which uses weakly-coupled low dimensional subsystems under the supervision of a dynamic coordinator program. This self-organizing adaptive control system was checked by a 5 MW research reactor. As an example the paper describes the experimental computer control system of the 5 MW WWR-SM research reactor, where the reactor power and outlet temperature have been controlled on the basis of the treated control concept since 1978. (author)

Preliminary Design of S-CO2 Brayton Cycle for KAIST Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Kim, Min Gil; Bae, Seong Jun; Lee, Jeong Ik

2013-01-01

This paper suggests a complete modular reactor with an innovative concept of reactor cooling by using a supercritical carbon dioxide directly. Authors propose the supercritical CO 2 Brayton cycle (S-CO 2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the core and PCU in one vessel for the full modularization. This study suggests a conceptual design of small modular reactor including PCU which is named as KAIST Micro Modular Reactor (MMR). As a part of ongoing research of conceptual design of KAIST MMR, preliminary design of power generation cycle was performed in this study. Since the targets of MMR are full modularization of a reactor system with S-CO 2 coolant, authors selected a simple recuperated S-CO 2 Brayton cycle as a power conversion system for KAIST MMR. The size of components of the S-CO 2 cycle is much smaller than existing helium Brayton cycle and steam Rankine cycle, and whole power conversion system can be contained with core and safety system in one containment vessel. From the investigation of the power conversion cycle, recompressing recuperated cycle showed higher efficiency than the simple recuperated cycle. However the volume of heat exchanger for recompressing cycle is too large so more space will be occupied by heat exchanger in the recompressing cycle than the simple recuperated cycle. Thus, authors consider that the simple recuperated cycle is more suitable for MMR. More research for the KAIST MMR will be followed in the future and detailed information of reactor core and safety system will be developed down the road. More refined cycle layout and design of turbomachinery and heat exchanger will be performed in the future study

A central European training course on reactor physics and kinetics - the 'Eugene Wigner Course' - Organisers view

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.; Matejka, K.; Sklenka, L.; Miglierini, M.; Sukods, C.

2004-01-01

Initiated by the 5th Framework Program of the European Commission, the European Nuclear Engineering Network (ENEN) is preparing the future European Nuclear Education schemes, degrees and requirements. To fully utilize the benefits of international cooperation and to promote the knowledge of students in nuclear engineering a 2.5 weeks course has been held, both in spring 2003 and 2004. The main emphasis of the course is to perform reactor physics and kinetics experiments on three different research- and training reactors in three different locations (Vienna, Prague, Budapest). The experimental work is preceded by theoretical lectures aiming to prepare the students for the experiments (Bratislava). The students' work will be evaluated, and upon success the students will get a certificate. The finally accepted credit (ECTS) value will be determined by the students' home university. The ENEN-recommended value is between 6 and 8 ECTS. The more detailed description of the course will be given in the full paper. (author)

Tourist Intensity in Capital Cities in Central Europe: Comparative Analysis of Tourism in Prague, Vienna and Budapest

OpenAIRE

Dumbrovská Veronika; Fialová Dana

2014-01-01

Urban tourism has become a significant phenomenon of tourism over the last decade. the importance of urban tourism has grown mainly due to the development of transport and information technologies. rapid advancement of low cost airlines and reduction of administrative barriers owing to the expansion of the schengen area caused not only the development of a number of urban destinations, including Prague, but also the growth of new source markets. this paper compares the development of urban to...

A comparison of male sex workers in Prague: Internet escorts versus men who work in specialized bars and clubs.

Science.gov (United States)

Bar-Johnson, Michael David; Weiss, Petr

2015-01-01

Prague, the Czech Republic, is a popular sex tourism destination where sex work is decriminalized and young men offer sexual services at low prices relative to countries in Western Europe. This quantitative survey aimed to identify some of the demographic characteristics of these young men and their experiences in the sex industry. Internet escorts (N = 20) and sex workers in bars and clubs (N = 20) completed the survey anonymously in spring 2011. The results showed that sex workers in clubs often had troubled pasts and were forced into sex work to survive. They also reported incidents of violence, serious alcohol and drug use, as well as frequent gambling. The larger group of sex workers in Prague is made up of Internet escorts who have backgrounds that are not atypical for the average Czech youth. They had fewer problems with drugs and alcohol but were twice as likely as sex workers in bars and clubs to be victims of violent crime. Plans for interventions to help those who would change their line of work, as well as the importance of sociocultural context in understanding sex workers, are discussed.

New digital control and power protection system of VR 1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Matejka, K.; Juoeickova, M.

2005-01-01

The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

Mukařovský´s Structuralism and Semiotics

Czech Academy of Sciences Publication Activity Database

Sládek, Ondřej

2016-01-01

Roč. 53, č. 2 (2016), s. 184-199 ISSN 0014-1291 Institutional support: RVO:68378068 Keywords : Mukařovský, Jan * structuralism * semiotics * Prague School Subject RIV: AJ - Letters, Mass-media, Audiovision

Development of first inhibitors for betaine-homocysteine S-methyltransferase 2

Czech Academy of Sciences Publication Activity Database

Mládková, J.; Vaněk, Václav; Elbert, Tomáš; Buděšínský, Miloš; Jiráček, Jiří

2012-01-01

Roč. 106, - (2012), s894-s894 ISSN 0009-2770. [EuCheMS Chemistry Congress /4./. 26.08.2012-30.08.2012, Prague] Institutional research plan: CEZ:AV0Z40550506 Keywords : enzyme catalysis * inhibitors * alkylation Subject RIV: CC - Organic Chemistry

The early Prague spring: Analysing the re-establishment of modernist aspects according to the example of three piano concertos by the “Prague group” of composers

Directory of Open Access Journals (Sweden)

Golubović Marija

2017-01-01

Full Text Available The interwar period brought about a number of modernist tendencies in the heterogeneous cultural context of the Kingdom of Yugoslavia, which is particularly salient in the works of the young composers belonging to the so-called “Prague group.” Having completed their studies, dozens of composers and conductors, including Ljubica Marić (1909-2003, Stanojlo Rajičić (1910-2000 and Milan Ristić (1908-1982 contributed to the establishment of the new movement in the conservative milieu of interwar Belgrade. After World War II, socialist realism became, in effect, the only approved style for the artists of the period. However, only a decade after the Tito-Stalin split, modernist tendencies reappeared fullblown in the output of Yugoslav composers. It is therefore of the greatest interest to analyse and present the way in which modernist music managed to find its way back to Yugoslav composers, performers and audiences in such a short period of time (the 1950s. To do so, we have chosen three piano concertos, written at the very beginning, in the middle, and at the very end of this period. This overview would not have been possible if we had analysed works belonging to other genres, as most had already been established in the pre-war period. However, it is also safe to conclude that the limitations on the Yugoslav scene were not imposed only by political authorities, but also by the conservative tastes of its audience and society, which were already in place before WWII.

Social Hygiene and Social Medicine in Interwar Czechoslovakia with the 13th District of the City of Prague as Its Laboratory

Directory of Open Access Journals (Sweden)

Mášová, Hana

2007-12-01

Full Text Available Health services in the newborn Czechoslovak Republic had to cope not only with the consequences of the First World War, but also with the new approaches to health policy. Modern social medicine challenged the conventional model of the so-called “police medicine” – the deep-rooted concept of Public Health in the Middle European countries. The necessity to resolve the contradiction between preventative and curative care led to the conclusion that the form and tasks of health centres, whether run by voluntary bodies or by the state/regional authorities, had to be reconsidered. The ambitious goal was to find new and more effective ways to cooperation between the various subjects of health assistance, consultancy, and – eventually - social aid. As an example of such an institution, providing both curative and preventive/consultant service, can be found in one of the most rapidly developing parts of the city of Prague – the new Czechoslovak capital. The “model district” in Prague XIII should have become the location utilizing the latest methods of social hygiene and healthcare organization, an enterprise of coordinated social work and health service. Only in the first years of its existence did it function in accordance with its aims; especially as a teaching arrangement - a tutorial service of the State Health Institute for the education of medical personnel, especially health and social nursing sisters, – and as a source of statistical research. The work of the voluntary organizations and health officers was successfully coordinated. But plans to reorganise the Health Office of the City of Prague, and to extend the system to other districts, faltered and eventually failed. Nevertheless, the principle was kept in mind by the later health care reformers, and survived.

Radiological protection of the staff during the decommissioning operations of the Romanian VVR-S research reactor

International Nuclear Information System (INIS)

Ene, D.C.

2002-01-01

Dose rate estimates for periods of 100 days and 6, 10, 25, 100 years after the shut down of the Romanian VVR-S reactor are presented in this paper for some foreseen decommissioning activities which include: i) cutting the water pipe in the pump room and the reactor sealing operations; ii) extracting reactor components; and iii) handling and dismantling the internal structures taken of from the reactor. For the reactor components extracted from the reactor, the considered calculation points were placed in the central plan of the items, on the surface and at distances from the surface which correspond to +0.2m, +1m, +2m, +8m, and +10m. Time dependence of the resulted dose rates are presented and discussed. Qualitative comparison with the measured values from other VVR-S reactors is done. The obtained results assist to develop working procedures that must be observed during the decommissioning activities. (author)

ANALÝZA PŘÍNOSU BĚŽECKÝCH ZÁVODŮ PRAGUE INTERNATIONAL MARATHON PRO CESTOVNÍ RUCH

OpenAIRE

Marková, Jana

2010-01-01

The thesis deals with the assessment of the sporting events impact. The main aim is to assess the impact of the running races organized by Prague International Marathon objectively and to verify the hypothesis that organizing sporting events can have a positive impact on tourism of the country. The theoretical part explains the basic approaches of sport tourism and deals with the typology of the sporting events. The practical part analyses the impact of the PIM running races on tourism. The a...

U.S. Non-proliferation policy and programs regarding use of high-enriched uranium in research reactors

International Nuclear Information System (INIS)

Lewis, R.A.

1993-01-01

Uranium enriched to 90-93%, supplied by the U.S., is now used in 141 research and test reactors in 35 countries around the world with a cumulative power of 1714 mw. Since of the order of 3 kg of 235 U is involved annually in fuel fabrication, fresh fuel transport and storage, reactor operation, and spent fuel cooling and return per megawatt of research reactor power, it is estimated that more than 5000 kg of very high-enriched uranium is handled each year to operate these reactors. Recent U.S. assessments have led to the tentative conclusion that in only approximately 11 of these reactors, generally those of highest power or power density, is the use of 90-93% enriched uranium currently a technical necessity. Universal use of the best state-of-the-art fuel technology would permit an estimated 90 of these reactors to use 20% enriched fuel, and estimated 40 others to use 45% enriched fuel, without significant performance degradation. If advanced research reactor fuel development programs currently under way in the U.S. and elsewhere are successful, it may, in fact, be possible to operate virtually all of these reactors on less than 20% enriched uranium in the longer term. The physical and economic practicality of these developmental fuels must, of course, await future assessments

THE EFFECT OF THE WINDOW-TO-WALL RATIO ON COOLING ENERGY USAGE AND COMFORT TEMPERATURE

Directory of Open Access Journals (Sweden)

Aris Budhiyanto

2017-12-01

Full Text Available This study presents an investigation of the effect of building envelope, especially glass facade buildings on cooling energy usage and thermal comfort. An office building was modeled with various window-to-wall ratio (WWR using panasap glass with SC=0.58 in order to analyze the effect of the WWR addition on cooling energy usage and comfort temperature. The result suggested that the average increase of the cooling energy usage is about 5.67% per 10% WWR addition, and of the operative temperature ranges from 0.350C to 0.560C per 10% WWR addition. Moreover, the building with above 20% WWR doesn’t provide comfort temperature.

Calibration of new I and C at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, Martin; Jurickova, Monika

2011-01-01

The paper describes a calibration of the new instrumentation and control (I and C) at the VR-1 training reactor in Prague. The I and C uses uncompensated fission chambers for the power measurement that operate in a pulse or a DC current and a Campbell regime, according to the reactor power. The pulse regime uses discrimination for the avoidance of gamma and noise influence of the measurement. The DC current regime employs a logarithmic amplifier to cover the whole reactor DC current power range with only one electronic circuit. The system computer calculates the real power from the logarithmic data. The Campbell regime is based on evaluation of the root mean square (RMS) value of the neutron noise. The calculated power from Campbell range is based on the square value of the RMS neutron noise data. All data for the power calculation are stored in computer flash memories. To set proper data there, it was necessary to carry out the calibration of the I and C. At first, the proper discrimination value was found while examining the spectrum of the neutron signal from the chamber. The constants for the DC current and Campbell calculations were determined from an independent reactor power measurement. The independent power measuring system that was used for the calibration was accomplished by a compensated current chamber with an electrometer. The calculated calibration constants were stored in the computer flash memories, and the calibrated system was again successfully compared with the independent power measuring system. Finally, proper gamma discrimination of the Campbell system was carefully checked.

Enabling IPv6 at FZU - WLCG Tier2 in Prague

International Nuclear Information System (INIS)

Kouba, Tomáš; Chudoba, Jiří; Eliáš, Marek

2014-01-01

The usage of the new IPv6 protocol in production is becoming reality in the HEP community and the Computing Centre of the Institute of Physics in Prague participates in many IPv6 related activities. Our contribution presents experience with monitoring in HEPiX distributed IPv6 testbed which includes 11 remote sites. We use Nagios to check availability of services and Smokeping for monitoring the network latency. Since it is not always trivial to setup DNS in a dual stack environment properly, we developed a Nagios plugin for checking whether a domain name is resolvable when using only IP protocol version 6 and only version 4. We will also present local area network monitoring and tuning related to IPv6 performance. One of the most important software for a grid site is a batch system for a job execution. We will present our experience with configuring and running Torque batch system in a dual stack environment. We also discuss the steps needed to run VO specific jobs in our IPv6 testbed.

Enabling IPv6 at FZU - WLCG Tier2 in Prague

Science.gov (United States)

Kouba, Tomáš; Chudoba, Jiří; Eliáš, Marek

2014-06-01

The usage of the new IPv6 protocol in production is becoming reality in the HEP community and the Computing Centre of the Institute of Physics in Prague participates in many IPv6 related activities. Our contribution presents experience with monitoring in HEPiX distributed IPv6 testbed which includes 11 remote sites. We use Nagios to check availability of services and Smokeping for monitoring the network latency. Since it is not always trivial to setup DNS in a dual stack environment properly, we developed a Nagios plugin for checking whether a domain name is resolvable when using only IP protocol version 6 and only version 4. We will also present local area network monitoring and tuning related to IPv6 performance. One of the most important software for a grid site is a batch system for a job execution. We will present our experience with configuring and running Torque batch system in a dual stack environment. We also discuss the steps needed to run VO specific jobs in our IPv6 testbed.

Decay mechanism of indoor porous opuka stone: a case study from the main altar located in the St. Vitus Cathedral, Prague (Czech Republic).

Czech Academy of Sciences Publication Activity Database

Přikryl, R.; Přikrylová, J.; Racek, M.; Weishauptová, Zuzana; Kreislová, K.

2017-01-01

Roč. 76, č. 7 (2017), č. článku 290. ISSN 1866-6280 Institutional support: RVO:67985891 Keywords : opuka stone * Prague * St.Vitus Cathedral * main altar * damage mechanisms * environmental monitoring Subject RIV: DB - Geology ; Mineralogy OBOR OECD: Geology Impact factor: 1.569, year: 2016

Experience in incorporation control of operation personnel of the Reinsberg nuclear power plant (1967 - 1973)

International Nuclear Information System (INIS)

Pul'khajm, K.F.; Klyuke, Kh.

1976-01-01

A theoretic estimate of fission products and corrosion nuclides in a WWR type reactor is given. The results of radiation monitoring of the personnel irradiation at the Reinsberg (DDR) nuclear power plant are presented. It is supposed that the intake of fission and corrosion products has been realized through inhalation. The equivalent doses of internal irradiation are calculated. The estimate of the results of radiation monitoring are presented [ru

In-Pile Assemblies for Investigation of Tritium Release from Li2TiO3 Lithium Ceramic

International Nuclear Information System (INIS)

Shestakov, V.; Tazhibayeva, I.; Kawamura, H.; Kenzhin, Y.; Kulsartov, T.; Chikhray, Y.; Kolbaenkov, A.; Arinkin, F.; Gizatulin, Sh.; Chakrov, P.

2005-01-01

The description of algorithm to design in-pipe experimental ampoule devices (IPAD) is presented here, including description of IPAD design for irradiation tests of highly enriched lithium ceramics at WWR-K reactor. The description of the system for registration of tritium release from ceramics during irradiation is presented as well. Typical curve of tritium release from the IPAD during irradiation under various temperatures of the samples is shown here

Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

International Nuclear Information System (INIS)

Wade, D.C.

1990-01-01

Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

Experimental validation of TASS/SMR-S critical flow model for the integral reactor SMART

Energy Technology Data Exchange (ETDEWEB)

Seo, Si Won; Ra, In Sik; Kim, Kun Yeup [ACT Co., Daejeon (Korea, Republic of); Chung, Young Jong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has a compact size and a relatively small power rating (330MWt) compared to a conventional reactor. Because new concepts are applied to SMART, an experimental and analytical validation is necessary for the safety evaluation of SMART. The analytical safety validation is being accomplished by a safety analysis code for an integral reactor, TASS/SMR-S developed by KAERI. TASS/SMR-S uses a lumped parameter one dimensional node and path modeling for the thermal hydraulic calculation and it uses point kinetics for the reactor power calculation. It has models for a general usage such as a core heat transfer model, a wall heat structure model, a critical flow model, component models, and it also has many SMART specific models such as an once through helical coiled steam generator model, and a condensate heat transfer model. To ensure that the TASS/SMR-S code has the calculation capability for the safety evaluation of SMART, the code should be validated for the specific models with the separate effect test experimental results. In this study, TASS/SMR-S critical flow model is evaluated as compared with SMD (Super Moby Dick) experiment

Quality control of baby food products on the basis of results obtained using the instrumental neutron-activation analysis technique

International Nuclear Information System (INIS)

Mukhammedov, S.; Khaydarov, A.; Pardaev, O.

2013-01-01

The purpose of this study was to use the instrumental neutron-activation analysis (INAA) to determine the elemental composition of some kind of imported baby food products (BFP) and to compare the results with the permissible contents. The nuclear reactor WWR-SM of INP has been used to develop INAA to study the mineral composition of some children's food products. The concentration of 26 trace elements, including Mg, Ca, Fe, Zn, etc. was found. The comparison of the results with regulation contents and the daily data on food needs have shown that the investigated group of BFP does not meet the requirements for all trace and macro elements composition. (authors)

Nuclear prehistory influence on transfer velocity of 54Mn impurity 'hot' atoms in irradiated metallic iron

International Nuclear Information System (INIS)

Alekseev, I.E.

2007-01-01

Influence of nuclear prehistory on transfer velocity of 54 Mn impurity 'hot'-atoms - got by different nuclear channels: 56 Fe(d, α), 54 Fe(n,p) in irradiated metallic iron - is studied. Irradiation of targets were carried out in U-120 accelerator (energy range 7.3/5.3 MeV, deuteron beam current makes up 5 μA). Mean density of thermal neutron (WWR-M reactor) makes up 8.6·10 13 neutron·cm -2 ·s -1 . It is shown, that transfer velocity of 54 Mn 'hot' atoms is defining by rate of radiation damage of targets in the irradiation process at that a key importance has a bombarding particles type applied for radioactive label getting

Eugene P. Wigner’s Visionary Contributions to Generations-I through IV Fission Reactors

Directory of Open Access Journals (Sweden)

Carré Frank

2014-01-01

Full Text Available Among Europe’s greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

International Nuclear Information System (INIS)

England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

1990-05-01

Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

International Nuclear Information System (INIS)

Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

1995-04-01

The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

Mathematical model use for evaluation of radioactivity spreading in nuclear power plant

International Nuclear Information System (INIS)

Kubik, I.; Gladki, Eh.; Yanchik, O.

1976-01-01

On the basis of knowledges of radioactive products behaviour and their spreading in nuclear power plant under normal and accident conditions a KOMPLEX program is developed in the FORTRAN 4 language, permitting to calculate the activity in separate parts of the nuclear power plant with WWR type reactor. The COMPLEX program includes the following subprograms: AZ - PRIM - for estimating active products in fuel, coolant, on the surfaces of fuel element cans and the primary circuit. The subprogram permits to estimate the coolant activity at the expense of fission fragments for 4 different leakage mechanisms: due to diffusion, considerable fuel element damage, contamination of fuel element can surface and fuel washout by coolant; KOR - the program for estimating active corrosion products; ACT - the program for estimating the activity of activation products; CONT - the program for estimating the activity in the nuclear power plant premises (protection envelop) and ventilating pipe. The desciption of the above subprograms is given. For testing of the mathematical model applicability and the possibilities of the corresponding programs the checking calculations for operating parameters of nuclear power plant with WWR type reactor were carried out. The calculation results obtained have shown the applicability of the model suggested and the corresponding programes for nuclear power plant under normal operation and accident conditions [ru

Actions for continued safe wet storage of spent nuclear fuel at VVR-S reactor in Bucharest-Magurele

International Nuclear Information System (INIS)

Isbasescu, M.; Zorliu, A.; Silviu-laurentiu, B.; Stefan, V. . E-mail address of corresponding author: mirifa@ifin.nipne.ro; Isbasescu, M.)

2005-01-01

The Romanian VVR-S research reactor is located 8 kilometers from Bucharest in the town of Magurele and was operated by the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH). The reactor first reached criticality in July 1957 and operated until December 1997 when it was permanently shutdown. The VVR - S reactor of IFIN has two repositories for spent fuel elements: (1) Cooling pool located in the reactor room; (2) Long-term repositories located outside the reactor building - SNFW (spent nuclear fuel warehouse). The major factors believed to influence the pitting of aluminium alloys are conductivity, pH, and bicarbonate, chloride, sulphate and oxygen content. Some of these parameters have been analyzed at SNFW-IFIN-HH. (author)

76 FR 44964 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on U.S...

Science.gov (United States)

2011-07-27

... Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on August 18, 2011, Room T-2B3, 11545 Rockville Pike... the meeting, if possible, so that appropriate arrangements can be made. Thirty-five hard copies of...

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

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

ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

International Nuclear Information System (INIS)

Belot, Clotilde

2013-01-01

The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

77 FR 74697 - Meeting of the ACRS, Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting

Science.gov (United States)

2012-12-17

..., Subcommittee on U.S. Evolutionary Power Reactor; Notice of Meeting The ACRS Subcommittee on U.S. Evolutionary Power Reactor (U.S. EPR) will hold a meeting on January 17, 2013, Room T-2B1, 11545 Rockville Pike... be made. Thirty-five hard copies of each presentation or handout should be provided to the DFO thirty...

Local bifurcation analysis in nuclear reactor dynamics by Sotomayor’s theorem

International Nuclear Information System (INIS)

Pirayesh, Behnam; Pazirandeh, Ali; Akbari, Monireh

2016-01-01

Highlights: • When the feedback reactivity is considered as a nonlinear function some complex behaviors may emerge in the system such as local bifurcation phenomenon. • The qualitative behaviors of a typical nuclear reactor near its equilibrium points have been studied analytically. • Comprehensive analytical bifurcation analyses presented in this paper are transcritical bifurcation, saddle- node bifurcation and pitchfork bifurcation. - Abstract: In this paper, a qualitative approach has been used to explore nuclear reactor behaviors with nonlinear feedback. First, a system of four dimensional ordinary differential equations governing the dynamics of a typical nuclear reactor is introduced. These four state variables are the relative power of the reactor, the relative concentration of delayed neutron precursors, the fuel temperature and the coolant temperature. Then, the qualitative behaviors of the dynamical system near its equilibria have been studied analytically by using local bifurcation theory and Sotomayor’s theorem. The results indicated that despite the uncertainty of the reactivity, we can analyze the qualitative behavior changes of the reactor from the bifurcation point of view. Notably, local bifurcations that were considered in this paper include transcritical bifurcation, saddle-node bifurcation and pitchfork bifurcation. The theoretical analysis showed that these three types of local bifurcations may occur in the four dimensional dynamical system. In addition, to confirm the analytical results the numerical simulations are given.

Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

International Nuclear Information System (INIS)

Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

2000-01-01

Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

A small research reactor for the 1980's

International Nuclear Information System (INIS)

Baglin, C.; Collis-Smith, J.A.; Mitchell, B.; Roskilly, T.

1978-01-01

In 1960, GEC together with Imperial College, designed and built the Consort reactor which is still in daily use at the London University Reactor Centre, Silwood Park. In 1977, GEC-REL chose the Consort reactor as a prototype for the development of a modern swimming pool research reactor, designed to meet the needs of countries or organisations starting in the field of Nuclear Technology. This paper outlines some of the topics which arose in the course of this project. (author)

U.S. Department of Energy operational experience with shipments of foreign research reactor spent nuclear fuel

International Nuclear Information System (INIS)

Messick, Charles E.; Massey, Charles D.; Mustin, Tracy P.

1998-01-01

On May 13, 1996, the U.S. Department of Energy issued a Record of Decision on a Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. The goal of the long-term policy is to recover enriched uranium exported from the United States, while giving foreign research reactor operators sufficient time to develop their own long-term solutions for storage and disposal of spent fuel. The spent fuel accepted by the U.S. DOE under the policy must be out of the research reactors by May 12, 2006 and returned to the United States by May 12, 2009. (author)

In-situ stripping of H{sub 2}S in gasoil hydrodesulphurization - reactor design considerations

Energy Technology Data Exchange (ETDEWEB)

Nava, J.A.O.; Krishna, R. [Amsterdam Univ., Dept. of Chemical Engineering, Amsterdam (Netherlands)

2004-02-01

In order to meet future diesel specifications the sulphur content of diesel would need to be reduced to below 50 ppm. This requirement would require improved reactor configurations. In this study we examine the benefits of counter-current contacting of gas oil with H{sub 2}, over conventional co-current contacting in a trickle bed hydrodesulphurization (HDS) reactor. In counter-current contacting, we achieve in-situ stripping of H{sub 2}S from the liquid phase; this is beneficial to the HDS kinetics. A comparison simulation study shows that counter-current contacting would require about 20% lower catalyst load than co-current contacting. However, counter-current contacting of gas and liquid phases in conventionally used HDS catalysts, of 1.5 mm sizes, is not possible due to flooding limitations. The catalysts need to be housed in special wire gauze envelopes as in the catalytic bales or KATAPAK-S configurations. A preliminary hardware design of a counter-current HDS reactor using catalytic bales was carried out in order to determine the technical feasibility. Using a realistic sulphur containing feedstock, the target of 50 ppm S content of desulphurized oil could be met in a reactor of reasonable dimensions. The study also underlines the need for accurate modelling of thermal effects during desulphurization. Our study also shows that interphase mass transfer is unlikely to be a limiting factor and there is a need to develop improved reactor configurations allowing for increased catalyst loading, at the expense of gas-liquid interfacial area. (Author)

Transformations of Opposition and Dissent in Prague and Brno in the Era of “the Normalization Regime”: Resistance to the Communist Regime between 1969 and 1989

Directory of Open Access Journals (Sweden)

Martina Miklová

2007-12-01

Full Text Available The article deals with various periods and changes relating to opposition and dissent in the time of the Czechoslovak “normalized regime”. The text is divided into four parts, where the author analyses questions concerning a term “the normalized regime”, b different periods and expressions acceptable for each phase, c activities produced by members of the resistance and d the forms of repression used against protagonists of opposing and dissident movements by the Communist regime and its secret police. The main objective of the article is to draw attention to specific features of Prague and Brno’s oppositional environment and to make connections between the situation in the CSSR and international events.

DEVELOPMENT OF AN ELECTRIC BICYCLE FOR A SHARING SYSTEM IN PRAGUE

Directory of Open Access Journals (Sweden)

William Deleenheer

2017-12-01

Full Text Available By means of a development of an e-bike sharing system the Electromobility Project wants to provide an alternative way of comfortable transportation for students and staff of the CTU, primarily to commute between different campuses. The research for this project contains at least three different fields of study, namely electric vehicle and docking station development, intelligent transport systems and management and economics of transportation and telecommunication. After briefly stating general requirements for the sharing system, this paper focuses on the development of the electric bicycle. First an ideal bike design is defined. Then necessary motor power and battery capacity are calculated by estimating characteristics of cycling in Prague. A prototype was developed by converting a normal bicycle to an electric bicycle. Being equipped with devices for e-bike monitoring, controlling and data recording for a post trip analysis, this prototype is also intended to have an educational value for future students in the project. Results consist of an electrical bicycle configuration that matches the requirements and a sketch of an ideal e-bike for this project.

The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

International Nuclear Information System (INIS)

Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

1992-01-01

A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

International Nuclear Information System (INIS)

Wright, A.E.; Dutt, D.S.; Harrison, L.J.

1990-01-01

Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

Proceedings of the International Association for Development of the Information Society (IADIS) International Conference on e-Learning (Prague, Czech Republic, July 23-26, 2013)

Science.gov (United States)

Nunes, Miguel Baptista, Ed.; McPherson, Maggie, Ed.

2013-01-01

These proceedings contain the papers of the International Conference e-Learning 2013, which was organised by the International Association for Development of the Information Society and is part of the Multi Conference on Computer Science and Information Systems (Prague, Czech Republic, July 23-26, 2013). The e-Learning 2013 conference aims to…

Neutron activation analysis of airborne inorganic pollutants

International Nuclear Information System (INIS)

Oerdoegh, M.; Kalman, L.

1975-01-01

The aim of the studies was to determine the radioactive and non-radioactive pollution of the air in the environment of the atomic reactor WWR-S of the Hungarian Academy of Sciences. Accordingly the investigations were carried out by two ways: the samples were analysed partly without previous irradiation and partly by neutron activation analysis. The use of several filter papers was studied and the most suitable filter paper (Whatman No. 41) was chosen for this investigation. The quantitative determination of more than twenty elements has been performed. First the natural activity of the aerosol samples taken periodically was measured and subsequently they have been analysed after irradiation in the atomic reactor for 1 minute, 1 hour, and 50 hours, respectively. A 45 cm 3 Ge/Li detector/ Nuclear Diodes/ was used in connection with a 1024 channel analyzer /type NTA-512B/. The analyses were made nondestructively, and the gamma-spectra were evaluated by computer. (K.A.)

Preliminary assessment of an S.G.H.W. type research reactor

International Nuclear Information System (INIS)

Bicevskis, A.; Chapman, A.G.; Hesse, E.W.

1970-08-01

A preliminary design study has been made of a research reactor, based on the enriched S.G.H.W.R. concept, to be used for power reactor fuel irradiation, isotope production, basic research, and training in nuclear technology. A reactor physics assessment established a core size which would allow uninterrupted operation for the required irradiation period consistent with low capital and operating costs. A design was selected with 24 channels, a D 2 O calandria diameter of 2.7 m and an overall core height of 4.0 m. The capital cost was estimated as $750,000 for the fuel and $1,600,000 for the moderator, the refuelling cost being $340,000 per annum. A thermal design study showed that the fission heat of 65 MW could be transmitted to pressurised light water at 200 lb/in 2 abs. and rejected to sea water in two conventional U-tube heat exchangers. The basic design is flexible and can be adapted to meet many special requirements. (author)

The regulation on commercial reactors and the management of high-level radioactive wastes in U.S

International Nuclear Information System (INIS)

Shimomura, Hidetsugu

2013-01-01

This article shows U.S. NRC's substantial and procedural regulations regarding commercial reactors and radioactive wastes. The commercial reactor's regulations are analyzed from an ensuring safety, and the radioactive waste' management is done from a locating a disposal site. (author)

Management of the radioactive waste resulting from the Romanian VVR-S research reactor decommissioning

International Nuclear Information System (INIS)

Ene, D.; Cepraga, D.G.

2002-01-01

The paper consists in a waste study of the Romanian VVR-S reactor which will be prepared for decommissioning operations after the permanent shutdown (23.12.1997). Calculations were carried out to determine the activity arising from neutron activation of structural materials inside the reactor, considering the design of the facility and its operating rules. To this end, the following method was used: i) Neutron flux distribution within the reactor was calculated using the DORT transport code, based on DLC23 shielding library relating to three cylindrical reference systems of the reactor structure: reactor core, horizontal tube and thermal column; ii) Calculation of the activity of each reactor component at different cooling times was performed by the ANITA2000 code, using the neutron flux, compositional data for each material and the power history of the reactor; iii) Unconditional clearance indexes for all material at various cooling times were calculated using the clearance levels defined in IAEA-TECDOC-855; iv) Total activities and masses by material type, within the waste category and for each decay time were calculated by summation of the data previously classified for each reactor component. The resulting activation inventory and waste masses, falling in IAEA defined waste categories are presented in the paper at periods of 100 days, and 6, 10, 25, and 50 years after reactor the shutdown. For some components of the reactor as: aluminum central vessel, the central iron shielding ring, the time behaviour of both the fin spatial activity distribution and the radionuclide contributions to the total activity are plotted in the paper. (author)

Integral Inherently Safe Light Water Reactor (I2S-LWR)

International Nuclear Information System (INIS)

Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

2017-01-01

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project ''Integral Inherently Safe Light Water Reactors (I 2 S-LWR)''. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

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

U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

International Nuclear Information System (INIS)

Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

2004-01-01

The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

Passive Decay Heat Removal System Options for S-CO2 Cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Moon, Jangsik; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

To achieve modularization of whole reactor system, Micro Modular Reactor (MMR) which has been being developed in KAIST took S-CO 2 Brayton power cycle. The S-CO 2 power cycle is suitable for SMR due to high cycle efficiency, simple layout, small turbine and small heat exchanger. These characteristics of S-CO 2 power cycle enable modular reactor system and make reduced system size. The reduced size and modular system motived MMR to have mobility by large trailer. Due to minimized on-site construction by modular system, MMR can be deployed in any electricity demand, even in isolated area. To achieve the objective, fully passive safety systems of MMR were designed to have high reliability when any offsite power is unavailable. In this research, the basic concept about MMR and Passive Decay Heat Removal (PDHR) system options for MMR are presented. LOCA, LOFA, LOHS and SBO are considered as DBAs of MMR. To cope with the DBAs, passive decay heat removal system is designed. Water cooled PDHR system shows simple layout, but has CCF with reactor systems and cannot cover all DBAs. On the other hand, air cooled PDHR system with two-phase closed thermosyphon shows high reliability due to minimized CCF and is able to cope with all DBAs. Therefore, the PDHR system of MMR will follows the air-cooled PDHR system and the air cooled system will be explored

Neutron field for activation experiments in horizontal channel of training reactor VR-1

Czech Academy of Sciences Publication Activity Database

Štefánik, Milan; Katovsky, K.; Vinš, M.; Šoltéš, J.; Závorka, L.

2014-01-01

Roč. 104, NOV (2014), s. 302-305 ISSN 0969-806X. [1st International Conference on Dosimetry and its Applications (ICDA). Prague, 23.6.2013-28.6.2013] R&D Projects: GA MŠk LG14004 Institutional support: RVO:61389005 Keywords : spectral index * neutron spectrometry * dosimetry-foils activation technique * irradiation channel * reaction rate * Gamma -spectroscopy Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.380, year: 2014

Safety evaluation report by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission for U.S. Energy Research and Development Administration Light Water Breeder Reactor. Special project No. 561

International Nuclear Information System (INIS)

1976-07-01

The Safety Evaluation Report is presented for the Light Water Breeder Reactor (LWBR). The LWBR core is to be installed in the Shippingport reactor at the Shippingport Atomic Power Station. The Safety Evaluation Report is the result of an NRC staff review of the LWBR Safety Analysis Report submitted by the Division of Naval Reactors, U. S. Energy Research and Development Administration. As a result of its review, the NRC staff has recommended that: (1) a diverse trip signal, such as containment high pressure, be included in a 2-out-of-3 logic for initiation of safety injection; (2) power be locked out from the pressurizer surge isolation valve during normal operation; and (3) a chlorine monitor be installed in the main control room

Experimental study on Response Parameters of Ni-rich NiTi Shape Memory Alloy during Wire Electric Discharge Machining

Science.gov (United States)

Bisaria, Himanshu; Shandilya, Pragya

2018-03-01

Nowadays NiTi SMAs are gaining more prominence due to their unique properties such as superelasticity, shape memory effect, high fatigue strength and many other enriched physical and mechanical properties. The current studies explore the effect of machining parameters namely, peak current (Ip), pulse off time (TOFF), and pulse on time (TON) on wire wear ratio (WWR), and dimensional deviation (DD) in WEDM. It was found that high discharge energy was mainly ascribed to high WWR and DD. The WWR and DD increased with the increase in pulse on time and peak current whereas high pulse off time was favourable for low WWR and DD.

Type specimens of centipedes (Myriapoda, Chilopoda in the National Museum, Prague (Czech Republic

Directory of Open Access Journals (Sweden)

Petr Dolejš

2015-06-01

Full Text Available The centipede collection in the National Museum in Prague contains type material of 16 taxa (14 species and two subspecies, of which 15 were described by Luděk J. Dobroruka and one by Karl W. Verhoeff: Allothereua wilsonae Dobroruka, 1979; Chinobius alenae Dobroruka, 1980; Lithobius corrigendus Dobroruka, 1988; L. creticus Dobroruka, 1977; L. erythrocephalus mohelensis Dobroruka, 1959; L. evae Dobroruka, 1958; L. magurensis Dobroruka, 1971; L. purkynei Dobroruka, 1957; L. tatricus Dobroruka, 1958; L. tatricus monounguis Dobroruka, 1958; Monotarsobius homolaci Dobroruka, 1971; M. krali Dobroruka, 1979; Pachymerium dilottiae Dobroruka, 1976; P. hanzaki Dobroruka, 1976; Scolopendra aztecorum Verhoeff, 1934 and Strigamia olympica Dobroruka, 1977. Of these 16 taxa, five were described from the Czech Republic, three from Slovakia and eight from other countries (Greece, Iraq, Kyrgyzstan, Mexico, Nepal, Russia and Uzbekistan. The eight taxa described from the Czech and Slovak Republics are now considered as junior synonyms but the eight taxa described from the other countries are still valid.

Twenty years of operation of the Radioisotope Department of the 3rd Medical Clinic, Faculty of General Medicine, Charles University in Prague

International Nuclear Information System (INIS)

Kapitola, J.

1983-01-01

Twenty years ago a radioisotope department was established in the old building of the 3rd Medical Clinic in Prague 2. The department is suitably placed and meets present requirements. It was set up as part of the 3rd Medical Clinic and of the Laboratory for endocrinology and Metabolism which gave it its main orientation and scope. Its present scope is much broader. In the twenty years since it was established 115,800 examinations were carried out, some 40 examination methods were introduced, 103 publications published, members of the department were co-authors of another 113 publications, they completed 11 research projects. The production of the department represents a substantial part of laboratory material especially in the diagnosis of endocrinopathy and metabolic disorders at the Clinic and is a significant part of the material of a number of research projects. The department has significantly contributed to the development of nuclear medicine in the Czechoslovak Socialist Republic in thyroid diagnosis, by the first introduction of radioimmunoassay methods, by the introduction of certain other special examination and laboratory methods and is currently taking part in the fulfilment of tasks given by the zoning of nuclear medicine in health care in Czechoslovakia in general and in Prague in particular. (author)

IPv6 testing and deployment at Prague Tier 2

Science.gov (United States)

Kouba, Tomáŝ; Chudoba, Jiří; Eliáŝ, Marek; Fiala, Lukáŝ

2012-12-01

Computing Center of the Institute of Physics in Prague provides computing and storage resources for various HEP experiments (D0, Atlas, Alice, Auger) and currently operates more than 300 worker nodes with more than 2500 cores and provides more than 2PB of disk space. Our site is limited to one C-sized block of IPv4 addresses, and hence we had to move most of our worker nodes behind the NAT. However this solution demands more difficult routing setup. We see the IPv6 deployment as a solution that provides less routing, more switching and therefore promises higher network throughput. The administrators of the Computing Center strive to configure and install all provided services automatically. For installation tasks we use PXE and kickstart, for network configuration we use DHCP and for software configuration we use CFEngine. Many hardware boxes are configured via specific web pages or telnet/ssh protocol provided by the box itself. All our services are monitored with several tools e.g. Nagios, Munin, Ganglia. We rely heavily on the SNMP protocol for hardware health monitoring. All these installation, configuration and monitoring tools must be tested before we can switch completely to IPv6 network stack. In this contribution we present the tests we have made, limitations we have faced and configuration decisions that we have made during IPv6 testing. We also present testbed built on virtual machines that was used for all the testing and evaluation.

IPv6 testing and deployment at Prague Tier 2

International Nuclear Information System (INIS)

Kouba, Tomáŝ; Chudoba, Jiří; Eliáŝ, Marek; Fiala, Lukáŝ

2012-01-01

Computing Center of the Institute of Physics in Prague provides computing and storage resources for various HEP experiments (D0, Atlas, Alice, Auger) and currently operates more than 300 worker nodes with more than 2500 cores and provides more than 2PB of disk space. Our site is limited to one C-sized block of IPv4 addresses, and hence we had to move most of our worker nodes behind the NAT. However this solution demands more difficult routing setup. We see the IPv6 deployment as a solution that provides less routing, more switching and therefore promises higher network throughput. The administrators of the Computing Center strive to configure and install all provided services automatically. For installation tasks we use PXE and kickstart, for network configuration we use DHCP and for software configuration we use CFEngine. Many hardware boxes are configured via specific web pages or telnet/ssh protocol provided by the box itself. All our services are monitored with several tools e.g. Nagios, Munin, Ganglia. We rely heavily on the SNMP protocol for hardware health monitoring. All these installation, configuration and monitoring tools must be tested before we can switch completely to IPv6 network stack. In this contribution we present the tests we have made, limitations we have faced and configuration decisions that we have made during IPv6 testing. We also present testbed built on virtual machines that was used for all the testing and evaluation.

U.S. Research Program to Support Advanced Reactors and Fuel Cycle Options

International Nuclear Information System (INIS)

Lyons, Peter

2013-01-01

• In recognition of possible future needs, the U.S. will perform R&D on advanced reactor and fuel cycle technologies that could dramatically improve nuclear energy safety and performance; • Multifaceted approach to support R&D: - National labs; - Universities; - Industry; - International partners

V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

International Nuclear Information System (INIS)

Teuchert, E.; Hansen, U.; Haas, K.A.

1980-03-01

V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

Reactor coolant pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Nuclear non-proliferation: the U.S. obligation to accept spent fuel from foreign research reactors

International Nuclear Information System (INIS)

Shapar, Howard K.; Egan, Joseph R.

1995-01-01

The U.S. Department of Energy (DOE) had a 35-year program for the sale and receipt (for reprocessing) of high-enriched research reactor fuel for foreign research reactors, executed pursuant to bilateral agreements with nuclear trading partners. In 1988, DOE abruptly let this program lapse, citing environmental obstacles. DOE promised to renew the program upon completion of an environmental review which was to take approximately six months. After three and a half years, an environmental assessment was finally produced.Over a year and half elapsed since publication of the assessment before DOE finally took action to renew the program. The paper sets forth the nuclear non-proliferation and related foreign policy considerations which support renewal of the program. It also summarized the contractual and other commitments made to foreign research reactors and foreign governments and aspects of U.S. environmental law as they apply to continuation of the program. (author)

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

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

Dry reloading and packaging of spent fuel at TRIGA MARK I reactor of Medical University Hanover (MHH), Germany

International Nuclear Information System (INIS)

Haferkamp, D.

2008-01-01

Between 1994 and 1998 the equipment for dry reloading of a research reactor was developed by Noell, which was funded by the German Federal Government and State of Saxonia. The task of this development programme was the design and delivery of an equipment able to load the spent fuel into the shipping casks in a dry mode for research reactors, where wet loading inside the storage pool is impossible. ALARA and infrastructure conditions had to be taken into consideration. Most of the research reactors of TRIGA MARK I type or WWR-SM have operating modes for handling of spent fuel inside the pond or for transfer of spent fuel from pond to dry/wet storage pools. On the other hand, most of them cannot handle heavy weighted shipping casks inside the reactor building because of the crane capacity, or inside water pool because of dimensions and weight of shipping casks. A typical licensed normal operating procedure for spent fuel in research reactors (TRIGA MARK I) is shown. Dry unloading procedure is described. Additionally to the normal operating procedures at the MHH research reactor the following steps were necessary: - dry packaging of spent fuel elements into the loading units (six packs) in order to minimise the transfer and loading steps between the pool and shipping cask; - transfer of spent fuel loading units from dry storage pool to the shipping cask (outside the reactor building) in a shielded transfer cask; - dry reloading of loading units, into the shipping casks outside the reactor building. The Dry Reloading Equipment implies the following 5 items: 1. loading units (six packs), which includes: - capacity up to six spent fuel elements; - criticality safe placement of spent fuel elements; - handling of several spent fuel elements in an aluminium loading unit. 2. Special Transfer Cask, which includes: - shielded housing with locks; - gripper inside housing; - hoist outside housing; - computer aided operation mode for loading and unloading. 3. Transfer Vehicle

Electromagnetic properties of REBaCuO superconducting tapes considered for magnets of fusion reactors.

Czech Academy of Sciences Publication Activity Database

Jirsa, Miloš; Rameš, Michal; Ďuran, Ivan; Entler, Slavomír; Melíšek, T.; Kováč, P.; Viererbl, L.

2017-01-01

Roč. 124, November (2017), s. 73-76 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 ; RVO:68378271 Keywords : Superconducting REBaCuO tapes * Magnetic hysteresis loops * Transport currents * Engineering currents * Angular dependence * Neutron irradiation Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (FZU-D) OBOR OECD: Nuclear related engineering; Nuclear related engineering (FZU-D) Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617304829

Demolice aneb jak si nedělat starosti s kulturní bombou v době politické prázdnoty

Czech Academy of Sciences Publication Activity Database

Mádl, Martin

2017-01-01

Roč. 47, č. 1 (2017), s. 26-27 ISSN 1213-4228 Institutional support: RVO:68378033 Keywords : czech architecture * monument preservation * Prague Subject RIV: AL - Art, Architecture , Cultural Heritage

Automation drying unit molybdenum-zirconium gel radioisotope production technetium-99M for nuclear medicine

International Nuclear Information System (INIS)

Chakrova, Y.; Khromushin, I.; Medvedeva, Z.; Fettsov, I.

2014-01-01

Full text : Since 2001 the Institute of Nuclear Physics of the Republic of Kazakhstan has began production of radiopharmaceutical based on technetium-99m from irradiated reactor WWR-K of natural molybdenum, which allows to obtain a solution of technetium-99m of the required quality and high volume activity. In 2013 an automated system is started, which is unique and urgent task is to develop algorithms and software in Python, as well as the manufacture of certain elements of technological systems for automated production

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

Effects of best practices to reduce sickness absenteeism in health care and welfare institutions : paper presented at the 10th EAWOP Congress in Prague (May 16-19, 2001)

NARCIS (Netherlands)

Vuuren, C.V. van; Gent, M.J. van; Frank, N.C.M.

2001-01-01

This paper was presented at the tenth Congress on Work and Organizational Psychology (EAWOP Congress) in Prague (May 16-19, 2001) and is based on a study among 1,600 employees in health care and welfare institutions, to find out what these institutions do about absenteeism and to see if whatever

Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

Science.gov (United States)

Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

2017-07-01

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

Energy Technology Data Exchange (ETDEWEB)

1976-05-01

Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

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

"Eurotrain for Training." Proceedings of a European Congress on Continuing Education and Training (4th, Berlin, Germany; Warsaw, Poland; Prague, Czechoslovakia; Budapest, Hungary; Vienna, Austria, October 5-9, 1992).

Science.gov (United States)

Wisser, Ulrike, Ed.; Grootings, Peter, Ed.

1992-01-01

A "travelling" congress was conducted in five European cities (Berlin, Warsaw, Prague, Budapest, and Vienna) to promote a mutual exchange of views between east and west. The participants stressed the growing European Community interest in current examples of cooperation with neighbors in central and eastern Europe. In addition to…

[Effectiveness of teaching gerontology and geriatrics in students of the 1st Faculty of Medicine, Charles University in Prague].

Science.gov (United States)

Mádlová, P; Neuwirth, J; Topinková, E

2006-01-01

Increasing number of seniors in the society requires more university-degree educated professionals--health care professionals, social care workers and managers with basic exposure to and knowledge of gerontology and geriatrics. The aim of our paper was to evaluate the effectiveness of undergraduate training of gerontology and geriatrics among students of the 1st Faculty of Medicine, Charles University in Prague. To get information about knowledge of medical students and students of ergotherapy and physiotherapy and about their attitudes towards senior citizens we conducted a survey using two anonymous questionnaires prepared in our department and piloted earlier. The survey ran during the academic year 2004/2005. Students completed identical questionnaires twice, first time before the start of the clinical rotation and second time after the training end (n=134). Evaluation of knowledge and attitudes confirmed that one to two weeks clinical rotation at Department of Geriatrics was effective and increased knowledge of students in the topic trained. The percentage of correct answers in all three evaluated training programmes increased after the completion of the clinical rotation and reached 83% and more. From 134 participating students, 54.5 % appreciated life experience and wisdom of seniors they met, 98.4 % of students were satisfied with the training programme and 67.2 % of students reported that after training they changed their attitude towards senior population. Our survey confirmed that clinical training in geriatric medicine at 1st Faculty of Medicine, Charles University in Prague, prepared in agreement with current European recommendations is sufficiently effective and well accepted by the students. Therefore we recommend introduction of formal geriatric training for students in all medical faculties in the Czech Republic.

Progress in the U.S. department of energy sponsored in-depth safety assessments of VVER and RBMK reactors

International Nuclear Information System (INIS)

Binder, J.L.; Petri, M.C.; Pasedag, W.F.

2001-01-01

Since the disastrous accident at Chernobyl Nuclear Power Plant Unit 4 in 1986, there has been international recognition of the safety concerns posed by the operation of 67 Soviet-designed commercial nuclear reactors. These reactors are operated in eight countries from the former Soviet Union and its former satellite states in Central and Eastern Europe. The majority of these plants are in the Russian Federation (30 units) and Ukraine (14 units). New plants are in various stages of construction. U.S. support to improve the safety of Soviet-designed reactors over the past decade has been intended to enhance operational safety, provide for risk-reduction measures, and enhance regulatory capability. The U.S. approach to improving the safety of Soviet-designed reactors has matured into a large multi-year program known as the Soviet-Designed Reactor Safety Program that is managed by the U.S. Department of Energy (US DOE). The mission of the program is to implement a self-sustaining nuclear safety improvement program that would lead to internationally accepted safety practices at the plants. Those practices would create a safety culture that would be reflected in the operation, regulation, and professional attitudes of the designers, operators, and regulators of the nuclear facilities. A key component of this larger program has been the Plant Safety Evaluation Program, which supports in-depth safety assessments of VVER and RBMK plants. (author)

Preliminary S-CO_2 Compressor Design for Micro Modular Reactor

International Nuclear Information System (INIS)

Lee, Jekyoung; Cho, Seong Kuk; Kim, Seong Gu; Lee, Jeong Ik

2016-01-01

Due to economic benefit of S-CO_2 Brayton cycle which is came from high efficiency and compactness, active research is currently conducted by various research groups and various approaches are suggested to take benefits of S-CO_2 Brayton cycle. KAIST research team also has been working on advanced concept for application of S-CO_2 Brayton cycle to nuclear system and Micro Modular Reactor (MMR) concept was suggested. The preliminary compressor design of S-CO_2 compressor for MMR system was carried out to observe feasibility of compressor design. Preliminary S-CO_2 compressor design for MMR system was successfully conducted and some issues are discovered from the design study. From the previous work done by Cho, conceptual design for MMR system was provided. Thus, further preliminary design should be carried out to obtain feasible S-CO_2 compressor design for MMR system. KAIST_TMD which is turbomachinery in-house code for real gases including S-CO_2 is continuously updated and currently it has 3D geometry construction and design optimization capability

Status and program of development of the fast breeder reactor system in the U.S

International Nuclear Information System (INIS)

Roberts, R.W.

1977-01-01

The U.S. Energy Research and Development Administration's highest priority breeder concept is the Liquid Metal Fast Breeder Reactor (LMFBR). Its objective is the development of a broad technological and engineering base with extensive utility and industrial involvement which will establish a timely capability for a competitive and commercial breeder industry. In addition to technological development, an integral part of the LMFBR program is the transfer of LMFBR technology to the nuclear industry and the parallel development of utility capability. An important indicator of the effectiveness of this program is the successful construction and operation of a number of LMFBR experimental and test reactors and associated non-nuclear test facilities such as those located at Santa Susana, California, the Argonne (III). National Laboratory and the Hanford Engineering Development Laboratory at Richland, Washington. A principal element of the U.S. fast breeder reactor program is the Fast Flux Test Facility (FFTF). The largest test reactor facility in the world, the FFTF will focus on the testing of fuels and material. The FFTF, being built at the Hanford Engineering Development Laboratory, is scheduled for completion in the late 1970's. The next step toward timely commercial breeder capability is to implement a program for a large-scale demonstration of this concept. A 380 megawatt electrical demonstration plant - designated the Clinch River Breeder Reactor Plant (CRBRP) - will be constructed and operated near Oak Ridge, Tennessee, under a cooperative arrangement of industrial contractors, utilities, and the Government. The completion date is 1983. The estimated cost of the CRBRP project is 1.95 billion dollars. This includes the cost of design, construction, related research and development, and five years of operation - the full timespan from the project's beginning in 1972 through completion. The Nation's principal electric utilities have strongly endorsed the

Selection of NPP for Kazakhstan

International Nuclear Information System (INIS)

Zhotabaev, Zh.R.

2003-01-01

Commercial NPP for Kazakhstan should to meet to several main requirements: 1). Safety operation (accident probability not more than 10 -6 1/p. year). 2). High efficiency > 40 %. 3). Possibility of use for high-temperature chemistry and hydrogen production. 4). Possibility for manufacturing of considerable part of equipment in Kazakhstan. 5). Possibility for fuel production and reprocessing in Kazakhstan. 6). Independence from existence of large water-supply sources. Comparative analysis of several NPP with different reactors (WWR-1000, Candu, BREST, VG-400; graphite molten salt reactor) shows that NPP with the graphite molten salt reactor meets to all above requirements, but hydrogen production it is possible by more complete 4-stage technology, since coolant temperature is 800 Deg. C. The principle advantage is possibility of manufacturing of main equipment and fuel in Kazakhstan that reduce the cost of NPP construction and operation

Spent nuclear fuel discharges from U.S. reactors 1994

International Nuclear Information System (INIS)

1996-02-01

Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year's report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs

Testing of methods for decontamination of stainless steels and carbon steels conformably to demountable equipment of nuclear power plant with WWR type reactor

International Nuclear Information System (INIS)

Dergunova, G.M.; Nazarov, V.K.; Ozolin, A.B.; Smirnov, L.M.; Stel'mashuk, V.P.; Yulikov, E.I.; Vlasov, I.N.

1978-01-01

Results are given of experiments on decontamination of stainless steel by the oxidation-reduction method and also results of decontamination of carbon steel by means of solutions based on oxalic acid, citric acid and phosphoric acid. Investigations of efficiency of oxidation-reduction treatment were done on samples of stainless steel cut from the pipeline of the primary coolant circuit of reactor. Comparison is given of efficiency of oxidation-reduction methods of contamination of stainless steel in the case of application of different compositions of decontaminating solutions. Dependences are given for decontamination completeness on duration of operations, on temperature and on ratio of volume of decontaminating solutions to surface are of the sample. For carbon steels parameters are given for decontamination process by means of oxalic, citric and phosphoric acid solutions. (I.T.) [ru

U.S. technology for mechanized/automated fabrication of fast reactor fuel

International Nuclear Information System (INIS)

Nyman, D.H.; Bennett, D.W.; Claudson, T.T.; Dahl, R.E.; Graham, R.A.; Keating, J.J.; Yatabe, J.M.

1978-01-01

The status of the U.S. fast reactor Fuel Fabrication Development Program is discussed. The objectives of the program are to develop and evaluate a high throughput pilot fuel fabrication line including close-coupled chemistry and wet scrap recycle operations. The goals of the program are to demonstrate by mechanized/automated and remote processes: reduced personnel exposure, enhanced safegurads/accountability, improved fuel performance, representative fabrication rates and reduced fuel costs

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

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

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

Aging management of PWR reactor internals in U.S. plants

International Nuclear Information System (INIS)

Amberge, K.J.; Demma, A.

2015-01-01

This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

U.S. Status of Fast Reactor Research and Technology

International Nuclear Information System (INIS)

Hill, Robert

2012-01-01

Summary: • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction) and safety: 1. System Integration and Concept Development; 2. Safety Technology; 3. Advanced Materials; 4. Ultrasonic Viewing; 5. Advanced Energy Conversion (Supercritical CO 2 Brayton cycle); 6. Reactor Simulation; 7. Nuclear Data; 8. Advanced Fuels. • Fast reactors have flexible capability for actinide management: – A wide variety of fuel cycle options are being considered; • International R&D collaboration pursued in Generation-IV and multilateral arrangements

Neutron diffraction radiation of solid solution of carbon and hydrogen in the α-titanium in the homogeneity domain

International Nuclear Information System (INIS)

Mirzaev, B.B.; Khidirov, I.; Mukhtarova, N.N.

2005-01-01

In the work by the neutron-graph the homogeneity domain of the introduction solid solution TiC x H y is determined. The sample neutron grams have been taken on the neutron diffractometer (λ=.1085 nm) installed at the thermal column of the WWR-SM reactor (INF AN RUz). For the phase analysis and estimation of solid solutions homogeneity the X-ray graph was used. X-ray grams were taken on the X-ray diffractometer DRON-3M with use of CuK α radiation (λ=0.015418 nm)

Reactor container

International Nuclear Information System (INIS)

Naruse, Yoshihiro.

1990-01-01

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

DIRECT ENERGY CONVERSION (DEC) FISSION REACTORS - A U.S. NERI PROJECT

International Nuclear Information System (INIS)

Beller, D.; Polansky, G.

2000-01-01

The direct conversion of the electrical energy of charged fission fragments was examined early in the nuclear reactor era, and the first theoretical treatment appeared in the literature in 1957. Most of the experiments conducted during the next ten years to investigate fission fragment direct energy conversion (DEC) were for understanding the nature and control of the charged particles. These experiments verified fundamental physics and identified a number of specific problem areas, but also demonstrated a number of technical challenges that limited DEC performance. Because DEC was insufficient for practical applications, by the late 1960s most R and D ceased in the US. Sporadic interest in the concept appears in the literature until this day, but there have been no recent programs to develop the technology. This has changed with the Nuclear Energy Research Initiative that was funded by the U.S. Congress in 1999. Most of the previous concepts were based on a fission electric cell known as a triode, where a central cathode is coated with a thin layer of nuclear fuel. A fission fragment that leaves the cathode with high kinetic energy and a large positive charge is decelerated as it approaches the anode by a charge differential of several million volts, it then deposits its charge in the anode after its kinetic energy is exhausted. Large numbers of low energy electrons leave the cathode with each fission fragment; they are suppressed by negatively biased on grid wires or by magnetic fields. Other concepts include magnetic collimators and quasi-direct magnetohydrodynamic generation (steady flow or pulsed). We present the basic principles of DEC fission reactors, review the previous research, discuss problem areas in detail and identify technological developments of the last 30 years relevant to overcoming these obstacles. A prognosis for future development of direct energy conversion fission reactors will be presented

Advanced boiling water reactors for the 90's and beyond

International Nuclear Information System (INIS)

Rao, A.S.; Sawyer, C.D.; Qurik, J.F.; McCandless, R.J.

1990-01-01

This paper discusses how the advanced boiling water reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability and reduced occupational exposure and radwaste. International cooperative efforts are also under way aimed at development of a simplified BWR employing natural circulation and passive safety systems. The SBWR conceptual design is complete. This BWR concept shows technical and economic promise. The SBWR program is aimed at providing a U.S. NRC certified design in an investor-ready state by 1995. With its short construction schedule, the 600 MWe SBWR will provide an option for commercial operation worldwide by the mid-to-late 1990s

Extensive utilization of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, Karel; Sklenka, Lubomir

2003-01-01

Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

Integral Inherently Safe Light Water Reactor (I²S-LWR)

Energy Technology Data Exchange (ETDEWEB)

Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

2017-10-02

This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I²S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

Engineering progress of CNS concept in Hanaro

International Nuclear Information System (INIS)

Choi, C.O.; Park, K.N.; Park, S.H.

1997-01-01

The Korea Atomic Energy research Institute (KAERI) strives to provide utilizing facilities on and around the Hanaro reactor in order to activate advanced researches by neutron application. As one of the facilities to be installed, the conceptual design work of CNS was started in 1996 with a project schedule of 5 years so that its installation work can be finished by the year 2000. And the major engineering targets of this CNS facility are established for a minimum physical interference with the present facilities of the Hanaro, a reach-out of very-high-gain factors in the cold neutron flux, a simplicity of the maintenance of the facility, and a safety in the operation of the facility as well as the reactor. For the conceptual design of Hanaro CNS, the experience of utilization and production of cold neutron at WWR-M reactor Gatchina, Russia has been used with that of elaborations for PIK reactor in design for neutron guide systems and instruments. (author)

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.

Nuclear chemistry on the Czech Technical University in Prague after introduction of structured study and foundation of the Centre for Radiochemistry and Radiation Chemistry

International Nuclear Information System (INIS)

John, J.

2007-01-01

In this presentation the author (head of the Centre for Radiochemistry and Radiation Chemistry) give a short review of history of the Department of Nuclear Chemistry and of the Centre for Radiochemistry and Radiation Chemistry of the Czech Technical University in Prague. Education in structured study in specialisation of nuclear chemistry in bachelor level, master level, as well as post-graduate study in nuclear chemistry with academic degree PhD. are realised. Some scientific results are presented

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

Science.gov (United States)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-02-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min-1, 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance.

K izučeniju kul'tury s keramikoj pražskogo tipa na territorii Moravii: formirovanije i osobennosti

Czech Academy of Sciences Publication Activity Database

Jelínková, Dagmar

-, č. 5 (2015), s. 117-150 ISSN 1608-9057 Institutional support: RVO:68081758 Keywords : Prague-type pottery culture * Moravia * dating * Cultural relations to the east and South-East * Influence of the Merovingian sphere Subject RIV: AC - Archeology, Anthropology, Ethnology

Application of S-CO_2 Cycle for Small Modular Reactor coupled with Desalination System

International Nuclear Information System (INIS)

Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik

2016-01-01

The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO_2 power cycle technology. The S-CO_2 Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO_2 cycles for SMART with desalination system is conducted. The simple recuperated S-CO_2 cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%

Recent progress in stellarator reactor conceptual design

International Nuclear Information System (INIS)

Miller, R.L.

1985-01-01

The Stellarator/Torsatron/Heliotron (S/T/H) class of toroidal magnetic fusion reactor designs continues to offer a distinct and in several ways superior approach to eventual commercial competitiveness. Although no major, integrated conceptual reactor design activity is presently underway, a number of international research efforts suggest avenues for the substantial improvement of the S/T/H reactor embodiment, which derive from recent experimental and theoretical progress and are responsive to current trends in fusion-reactor projection to set the stage for a third generation of designs. Recent S/T/H reactor design activity is reviewed and the impact of the changing technical and programmatic context on the direction of future S/T/H reactor design studies is outlined

Work Breakdown Structure and Work Packages for Decommissioning the Nuclear Research Reactor VVR-S Magurele-Bucharest

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-06-15

The research reactor type VVR-S (tank type, water cooled, moderator and reflector, thermal power 2 MW, thermal energy 9.52 GWd) was put into service in July 1957, and in December 1997, was shut down. In 2002, the Romanian Government decided to put the research reactor into a permanent shutdown condition in order to start decommissioning. This nuclear facility had been used in nuclear research and radioisotope production for 40 years without any events, incidents or accidents. At the same site, in the immediate vicinity of the research reactor, there are many other nuclear facilities: a radioactive waste treatment plant, a tandem Van de Graaff heavy ion accelerator, a cyclotron, an industrial irradiator and a radioisotope production centre.

Application of LiF for determining the gamma-radiation characteristics of the shut-down reactor

Energy Technology Data Exchange (ETDEWEB)

Ibragimova, E M; Musaeva, M A; Ashrapov, U T; Kalanov, M U; Muminov, M I [Inst. of Nuclear Physics, Tashkent (Uzbekistan)

2005-07-01

Full text: The power of {sup 60}Co {approx}1.25 MeV gamma-radiation source at the INP AS RUz is limited by 8 Gy/s, which does not satisfy several tasks of material science now. Therefore, we were first to suggest the irradiation of materials with gamma-rays of 0.1-7 MeV, which are emitted by the uranium fission products ({sup 41}Ar, {sup 135}Xe, {sup 125}Xe, {sup 125}I,{sup 137}Cs, {sup 134}Cs, {sup 144}Ce, {sup 95}Zr, {sup 140}Ba, {sup 140}La, {sup 99}Mo, {sup 60}Co) and {sup l6}N, {sup 24}Na, {sup 28}Al radio-nuclides in water during prophylactic shut-downs of our nuclear reactor WWR-SM. The gamma-dose rate kinetics was monitored with the ion current in ionization chambers KNK-53M fixed outside the reactor core from the stop-moment. The current kinetics comprised 4 steps with a high reproducibility at 2 and 0.5 {mu}A, then 50 and 10 nA, each lasting for 1,10, 40 and up to 200 hours, according to the isotope life-times. LiF crystal is known as a thermal luminescence dosimeter of mixed radiations up to 100 Gy. Yet in this work the absorbed gamma-energy dose D{sub {gamma}} was determined by accumulation of the known stable structure defects in thin cleaved LiF crystals: by induced optical absorption and luminescence of F- and M-centers. The samples were irradiated in Al-containers filled with water to keep the temperature of {approx}40 deg. C in the time range from 30 minutes to 150 hours. Optical absorption spectra were registered at spectrometer Specord M-40. Then the induced color center concentration was calculated by the Smakula relation, which is proportional to the absorbed dose D{gamma}. For a better reliability the photoluminescence center content was also determined. Selecting comparable close intensities of the induced absorption and luminescence bands obtained after irradiations of LiF references in the certified {sup 60}Co gamma-sources of the known gamma fluxes 0.7 and 7.5 Gy/s, the gamma-radiation intensity of the shut-down reactor was estimated in

Application of LiF for determining the gamma-radiation characteristics of the shut-down reactor

International Nuclear Information System (INIS)

Ibragimova, E.M.; Musaeva, M.A.; Ashrapov, U.T.; Kalanov, M.U.; Muminov, M.I.

2005-01-01

Full text: The power of 60 Co ∼1.25 MeV gamma-radiation source at the INP AS RUz is limited by 8 Gy/s, which does not satisfy several tasks of material science now. Therefore, we were first to suggest the irradiation of materials with gamma-rays of 0.1-7 MeV, which are emitted by the uranium fission products ( 41 Ar, 135 Xe, 125 Xe, 125 I, 137 Cs, 134 Cs, 144 Ce, 95 Zr, 140 Ba, 140 La, 99 Mo, 60 Co) and l6 N, 24 Na, 28 Al radio-nuclides in water during prophylactic shut-downs of our nuclear reactor WWR-SM. The gamma-dose rate kinetics was monitored with the ion current in ionization chambers KNK-53M fixed outside the reactor core from the stop-moment. The current kinetics comprised 4 steps with a high reproducibility at 2 and 0.5 μA, then 50 and 10 nA, each lasting for 1,10, 40 and up to 200 hours, according to the isotope life-times. LiF crystal is known as a thermal luminescence dosimeter of mixed radiations up to 100 Gy. Yet in this work the absorbed gamma-energy dose D γ was determined by accumulation of the known stable structure defects in thin cleaved LiF crystals: by induced optical absorption and luminescence of F- and M-centers. The samples were irradiated in Al-containers filled with water to keep the temperature of ∼40 deg. C in the time range from 30 minutes to 150 hours. Optical absorption spectra were registered at spectrometer Specord M-40. Then the induced color center concentration was calculated by the Smakula relation, which is proportional to the absorbed dose Dγ. For a better reliability the photoluminescence center content was also determined. Selecting comparable close intensities of the induced absorption and luminescence bands obtained after irradiations of LiF references in the certified 60 Co gamma-sources of the known gamma fluxes 0.7 and 7.5 Gy/s, the gamma-radiation intensity of the shut-down reactor was estimated in correlation with the ion current as 10 nA = 15 Gy/s. At short times of irradiation the linear dose dependence

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

Characterization of Indoor and Outdoor Aerosols in a Suburban Area of Prague

International Nuclear Information System (INIS)

Smolik, J.; Dohanyosova, P.; Schwarz, J.; Zdimal, V.; Lazaridis, M.

2008-01-01

The mass, ionic and elemental size distributions of particulate matter (PM) measured indoors and outdoors in an apartment situated in a north-westward suburb of Prague are presented. The PM samples were collected by two Berner type low pressure impactors separating particles into 10 size fractions from 26 nm to 10 μm and were further analyzed by ion chromatography (IC) and proton induced X-ray emission (PIXE). Temperature, pressure and relative humidity were measured both indoors and outdoors parallel to PM sampling. The indoor and outdoor PM dynamics were recorded by two scanning mobility particle sizers (SMPS) and an aerodynamic particle sizer (APS). Finally, the ventilation rate was determined by a radon technique. Ion chromatography showed that the major inorganic components of the fine particle mode are sulfate, nitrate, and ammonium with very low indoor nitrate concentration. Crustal elements (Al, Si, Ca, Ti, Mn, and Fe) were associated with the coarse aerosol mode. The presence of people increased the mass concentration of coarse particles, whereas cooking, smoking, and burning of incense and candles contributed predominantly to the fine particle mode. Smoking and the burning of incense also increased the concentration of potassium, bromine and chlorine content in fine particles

The nineteen theses of Georges Florovsky’s dissertation “The hertsen’s philosophy of history”

Directory of Open Access Journals (Sweden)

Beiker Matfei

2013-06-01

Full Text Available Paul Gavrilyuk recently reconstructed Georges Florovsky’s dissertation, “Hertsen’s Philosophy of History”, the complete text of which was previously considered lost. According to the report of Alexander Izgoev, G.V. Florovsky presented the summary of his dissertation in 19 theses during his dissertation defense in Prague on 3 June 1923. The typescript of these 19 theses, bearing the title «The Dissertation’s Theses of Privatdozent Georges Florovsky: «The Hertsen’s Philosophy of History»» was recently discovered by Matthew Baker in the Andrew Blane Papers. This document is published here for the first time. The 19 theses support the Gavrilyuk’s hypothesis that the June 1923 redaction of the dissertation consisted of an introduction, three chapters and a conclusion.

Update on reactors and reactor instruments in Asia

Science.gov (United States)

Rao, K. R.

1991-10-01

The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

Update on reactors and reactor instruments in Asia

International Nuclear Information System (INIS)

Rao, K.R.

1991-01-01

The 1980s have seen the commissioning of several medium flux (∝10 14 neutrons/cm 2 s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high-Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand. (orig.)

Mental stress and hypertension, an evolutionary framework: some historical perspectives of the 1960 World Health Organization Prague Hypertension Meeting.

Science.gov (United States)

Kuchel, Otto

2003-03-01

Emotional stress acutely and repetitively causing blood pressure increase or aggravating existing hypertension is usually not reflected by norepinephrine and epinephrine increase but by a sudden rise of dopamine, the third "defensive" catecholamine coping with the damaging neuropsychological and cardiovascular actions of the first two. This double-edged sympathetic response to emotional stress evolves during human lifespan and long-term evolution of hypertension. In the course of philogenesis it carries a potential mismatch between the normal physiology of the human dopaminergic system and current environmental (emotional particularly) conditions in industrialized countries. This offers a rational support to a mental stress-cardiovascular diseases relationship proposed 40 years ago in a WHO report which followed a memorable 1960 Prague Hypertension Meeting.

Internet accessible hot cell with gamma spectroscopy at the Missouri S and T nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Grant, Edwin [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States); Mueller, Gary, E-mail: gmueller@mst.edu [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States); Castano, Carlos; Usman, Shoaib; Kumar, Arvind [Nuclear Engineering, Missouri University of Science and Technology, 203 Fulton Hall, 300 W. 13th St., Rolla, MO 65409 (United States)

2011-08-15

Highlights: > A dual-chambered internet-accessible heavily shielded facility has been built. > The facility allows distance users to analyze neutron irradiated samples remotely. > The Missouri S and T system uses computer automation with user feedback. > The system can analyze multiple samples and assist several researchers concurrently. - Abstract: A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S and T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently available for irradiation and analysis of samples. The facility allows authorized distance users engaged in collaborative activities with Missouri S and T to remotely manipulate and analyze neutron irradiated samples. The system consists of two shielded compartments, one for multiple sample storage, and the other dedicated exclusively for radiation measurements and spectroscopy. The second chamber has multiple detector ports, with graded shielding, and has the capability to support gamma spectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connected though a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based system complements the MSTR's current bare pneumatic tube (BPT) and cadmium lined pneumatic tube (CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g, irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals who do not have on-site access to a nuclear reactor facility to remotely participate in research and educational activities.

Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

International Nuclear Information System (INIS)

Pomorski, Michal; Mer-Calfati, Christine; Foulon, Francois; Sklenka, Lubomir; Rataj, Jan; Bily, Tomas

2015-01-01

Diamond exhibits a combination of properties which makes it attractive for neutron detection in hostile conditions. In the particular case of detection in a nuclear reactor, it is resilient to radiation, exhibits a natural low sensitivity to gamma rays, and its small size (as compared with that of gas ionisation chambers) enables fluency monitoring with a high position resolution. We report here on the use of synthetic CVD diamond as a solid state micro-fission chamber with U-235 converting material for in-core thermal neutron monitoring. Two types of thin diamond detectors were developed for this application. The first type of detector is fabricated using thin diamond membrane obtained by etching low-cost commercially available single crystal CVD intrinsic diamond, so called 'optical grade' material. Starting from a few hundred of micrometre thick samples, the sample is sliced with a laser and then plasma etched down to a few tenths of micrometre. Here we report the result obtained with a 17 μm thick device. The detection surface of this detector is equal to 1 mm 2 . Detectors with surfaces up to 1 cm 2 can be fabricated with this technique. The second type of detector is fabricated by growing successively two thin films of diamond, by the microwave enhanced chemical vapour deposition technique, on HPHT single crystal diamond. A first, a film of boron doped (p+) single crystal diamond, a few microns thick, is deposited. Then a second film of intrinsic diamond with a thickness of a few tens of microns is deposited. This results in a P doped, Intrinsic, Metal structure (PIM) structure in which the intrinsic volume id the active part of the detector. Here we report the results obtained with a 20 μm thick intrinsic whose detection surface is equal to 0.5 mm 2 , with the possibility to enlarge the surface of the detector up to 1 cm 2 . These two types of detector were tested at the VR-1 research reactor at the Czech Technical University in Prague. The

Successive and large-scale synthesis of InP/ZnS quantum dots in a hybrid reactor and their application to white LEDs

International Nuclear Information System (INIS)

Kim, Kyungnam; Jeong, Sohee; Woo, Ju Yeon; Han, Chang-Soo

2012-01-01

We report successive and large-scale synthesis of InP/ZnS core/shell nanocrystal quantum dots (QDs) using a customized hybrid flow reactor, which is based on serial combination of a batch-type mixer and a flow-type furnace. InP cores and InP/ZnS core/shell QDs were successively synthesized in the hybrid reactor in a simple one-step process. In this reactor, the flow rate of the solutions was typically 1 ml min −1 , 100 times larger than that of conventional microfluidic reactors. In order to synthesize high-quality InP/ZnS QDs, we controlled both the flow rate and the crystal growth temperature. Finally, we obtained high-quality InP/ZnS QDs in colors from bluish green to red, and we demonstrated that these core/shell QDs could be incorporated into white-light-emitting diode (LED) devices to improve color rendering performance. (paper)

U.S. NRC training for research and training reactor inspectors

International Nuclear Information System (INIS)

Sandquist, G.M.; Kunze, J.F.

2011-01-01

Currently, a large number of license activities (Early Site Permits, Combined Operating License, reactor certifications, etc.), are pending for review before the United States Nuclear Regulatory Commission (US NRC). Much of the senior staff at the NRC is now committed to these review and licensing actions. To address this additional workload, the NRC has recruited a large number of new Regulatory Staff for dealing with these and other regulatory actions such as the US Fleet of Research and Test Reactors (RTRs). These reactors pose unusual demands on Regulatory Staff since the US Fleet of RTRs, although few (32 Licensed RTRs as of 2010), they represent a broad range of reactor types, operations, and research and training aspects that nuclear reactor power plants (such as the 104 LWRs) do not pose. The US NRC must inspect and regulate all these entities. This paper addresses selected training topics and regulatory activities provided US NRC Inspectors for US RTRs. (author)

K aktivnímu životu seniorů. Dům s pečovatelskou službou v Úvalech u Prahy

Czech Academy of Sciences Publication Activity Database

Svobodová, Markéta

2003-01-01

Roč. 49, č. 8 (2003), s. 17 ISSN 0862-7010 Institutional research plan: CEZ:AV0Z8033913 Keywords : Czech architecture * home for the elderly * Prague Subject RIV: AL - Art, Architecture, Cultural Heritage

Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

International Nuclear Information System (INIS)

Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

2017-01-01

Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

Fourth Generation Reactor Concepts

International Nuclear Information System (INIS)

Furtek, A.

2008-01-01

Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

Reactor building

International Nuclear Information System (INIS)

Ebata, Sakae.

1990-01-01

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

Participation in the U.S. Department of Energy Reactor Sharing Program

International Nuclear Information System (INIS)

Mulder, R. U.; Benneche, P. E.; Hosticka, B.

1998-01-01

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

HOW MUCH DOES A MINUTE OF COMMUTING TIME COST? AN EXAMINATION OF PROPERTY PRICES IN RELATION TO DISTANCE TO THE CITY CENTER IN PRAGUE, CZECH REPUBLIC

Directory of Open Access Journals (Sweden)

Martin Lukavec

2017-12-01

Full Text Available This paper sets out to explore the strength of the relationship between the proximity of a property to the city center and its price. Buyers are willing to pay extra for apartments or houses closer to the city center, but the extent of this willingness remains largely unexplored. Our research question is: How much does a minute of commuting time influence the price of an apartment in Prague? In other words, with every minute of commuting time, how much more is paid for a house or an apartment closer to the central business district (CBD? Our analysis has found that on average, every minute of commuting time closer to the city center corresponds to an additional cost of CZK 43,390.45 for an average sized apartment in Prague. A regression analysis is graphically plotted in the Chart 1. We have also found that this relationship changes according to distance from the city center. For a commuting time of 1-20 minutes to the city center, the price increase is the highest: CZK 259,466.18 per minute. However, this figure is only CZK 55,809.01 for the interval of 21-40 minutes, and CZK 33,924.29 per minute for the interval of 41-55 minutes.

Reactor Coolant Pump seal issues and their applicability to new reactor designs

International Nuclear Information System (INIS)

Ruger, C.J.; Higgins, J.C.

1993-01-01

Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

Reactor building

International Nuclear Information System (INIS)

Maruyama, Toru; Murata, Ritsuko.

1996-01-01

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

Development and applications of reactor noise analysis at Ontario Hydro`s CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O [Ontario Hydro, Toronto, ON (Canada); Tulett, M V [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station

1996-12-31

In 1992 a program was initiated to establish reactor noise analysis as a practical tool for plant performance monitoring and system diagnostics in Ontario Hydro`s CANDU reactors. Since then, various CANDU-specific noise analysis applications have been developed and validated. The noise-based statistical techniques are being successfully applied as powerful troubleshooting and diagnostic tools to a wide variety of actual operational I and C problems. The dynamic characteristics of critical plant components, instrumentation and processes are monitored on a regular basis. Recent applications of noise analysis include (1) validating the dynamics of in-core flux detectors (ICFDS) and ion chambers, (2) estimating the prompt fraction ICFDs in noise measurements at full power and in power rundown tests, (3) identifying the cause of excessive signal fluctuations in certain flux detectors, (4) validating the dynamic coupling between liquid zone control signals, (5) detecting and monitoring mechanical vibrations of detector tubes induced by moderator flow, (6) estimating the dynamics and response time of RTD (Resistance Temperature Detector) temperature signals, (7) isolating the cause of RTD signal anomalies, (8) investigating the source of abnormal flow signal behaviour, (9) estimating the overall response time of flow and pressure signals, (10) detecting coolant boiling in fully instrumented fuel channels, (11) monitoring moderator circulation via temperature noise, and (12) predicting the performance of shut-off rods. Some of these applications are performed on an as-needed basis. The noise analysis program, in the Pickering-B station alone, has saved Ontario Hydro millions of dollars during its first three years. The results of the noise analysis program have been also reviewed by the regulator (Atomic Energy Control Board of Canada) with favorable results. The AECB have expressed interest in Ontario Hydro further exploiting the use of noise analysis technology. (author

Overview of U.S. Fast Reactor Technology Program

International Nuclear Information System (INIS)

Hill, Robert

2013-01-01

• Concept development studies guide R&D tasks by evaluating system impact for broad variety of technology options: – Small-scale facilities for R&D on key technology; – No near-term plan for demonstration reactor. • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction): – Advanced Structural Materials; – Advanced Energy Conversion; – Advanced Modeling and Simulation. • Other R&D is conducted to address known technology challenges: – Safety and Licensing; – Fuels Development; – Undersodium Viewing

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

Seven years of operation of the U. S. geological survey TRIGA reactor

International Nuclear Information System (INIS)

Kraker, Pat

1976-01-01

February 1976 marks 7 years of operation of the U. S. Geological Survey TRIGA Reactor (GSTR) facility. In these 7 years we have generated more than 5800 MWH's of thermal energy and irradiated more than 47,000 samples for experimenters from the Survey, universities, and other Governmental agencies. Several mechanical and electrical components have required attention. Changes to the technical specifications have included one minor wording change involving the evacuation alarm, a reevaluation of the measurement of argon-41 concentrations, a revision concerning transient-rod maintenance, and a reduction in the frequency of fuel-element measurements. To improve physical security we have increased building security, installed an intrusion alarm, and, most recently, expanded the boundaries of the facility within the building to provide better control access. There also have been major changes to our operating procedures and the initiation of a reactor-operator requalification program. (author)

Industrial structure at research reactor suppliers

International Nuclear Information System (INIS)

Roegler, H.-J.; Bogusch, E.; Friebe, T.

2001-01-01

Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

ZEEP: Canada's first nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Green, R.E.; Okazaki, A. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2015-09-15

In 1905 Albert Einstein published his historic paper on special relativity, which contained the equation E=mc 2. The significance of this mass-energy relationship became evident with the discovery of nuclear fission in 1939, when it was realized that large amounts of energy would be released in a fission chain reaction. Canadian scientists were involved in this field from the beginning and their efforts resulted in the startup in September 1945 of the ZEEP reactor at Chalk River, the first reactor to go critical outside the USA. In this paper we recall some of the events that led to the construction of ZEEP, and describe the role it played in the development of the Canadian nuclear energy program. (author)

National report from Czech Republic

International Nuclear Information System (INIS)

Kurka, J.

1995-01-01

The project of the Temelin NPP instrumentation and control systems and reactor core fuel replacement is outlined. Current status in development of operator support system and the activities in EGU Prague, plc. under CRP contract with this regard is discussed

Direct energy conversion in fission reactors: A U.S. NERI project

International Nuclear Information System (INIS)

Slutz, Stephen A.; Seidel, David B.; Polansky, Gary F.; Rochau, Gary E.; Lipinski, Ronald J.; Besenbruch, G.; Brown, L.C.; Parish, T.A.; Anghaie, S.; Beller, D.E.

2000-01-01

In principle, the energy released by a fission can be converted directly into electricity by using the charged fission fragments. The first theoretical treatment of direct energy conversion (DEC) appeared in the literature in 1957. Experiments were conducted over the next ten years, which identified a number of problem areas. Research declined by the late 1960's due to technical challenges that limited performance. Under the Nuclear Energy Research Initiative the authors are determining if these technical challenges can be overcome with todays technology. The authors present the basic principles of DEC reactors, review previous research, discuss problem areas in detail, and identify technological developments of the last 30 years that can overcome these obstacles. As an example, the fission electric cell must be insulated to avoid electrons crossing the cell. This insulation could be provided by a magnetic field as attempted in the early experiments. However, from work on magnetically insulated ion diodes they know how to significantly improve the field geometry. Finally, a prognosis for future development of DEC reactors will be presented

Notes on Women Who Rock: Making Scenes, Building Communities: Participatory Research, Community Engagement, and Archival Practice

Directory of Open Access Journals (Sweden)

Michelle Habell-Pallán

2014-07-01

Full Text Available Since 2011, Women Who Rock (WWR has brought together scholars, archivists, musicians, media-makers, performers, artists, and activists to explore the role of women and popular music in the creation of cultural scenes and social justice movements in the Americas and beyond. The project promotes generative dialogue and documentation by “encompassing several interwoven components: project-based coursework at the graduate and undergraduate levels; an annual participant-driven conference and film festival; and an oral history archive hosted by the University of Washington Libraries Digital Initiatives Program that ties the various components together” (Bartha 8. In our courses, programming, and archive, we examine the politics of performance, social identity, and material access in music scenes, cultures, and industries. Performance studies scholar Daphne Brooks argues that the “confluence of cultural studies, rock studies, and third wave feminist critical studies makes it possible now more than ever to continue to critique and re-interrogate the form and content of popular music histories” (58. WWR implements this approach, asking how particular stories of popular music determine a performer, band, or scene’s “legendary” status or excision from the official annals of memory. WWR reshapes conventional understandings of popular music studies by initiating collective methods of participatory research, as well as community collaboration and dialogue. By way of WWR, we seek to transform traditional models of popular music studies, instigating new convergences between academic disciplines and critical approaches that create alternative histories and new forms of knowledge.

Internet accessible hot cell with gamma spectroscopy at the Missouri S and T nuclear reactor

International Nuclear Information System (INIS)

Grant, Edwin; Mueller, Gary; Castano, Carlos; Usman, Shoaib; Kumar, Arvind

2011-01-01

Highlights: → A dual-chambered internet-accessible heavily shielded facility has been built. → The facility allows distance users to analyze neutron irradiated samples remotely. → The Missouri S and T system uses computer automation with user feedback. → The system can analyze multiple samples and assist several researchers concurrently. - Abstract: A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S and T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently available for irradiation and analysis of samples. The facility allows authorized distance users engaged in collaborative activities with Missouri S and T to remotely manipulate and analyze neutron irradiated samples. The system consists of two shielded compartments, one for multiple sample storage, and the other dedicated exclusively for radiation measurements and spectroscopy. The second chamber has multiple detector ports, with graded shielding, and has the capability to support gamma spectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connected though a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based system complements the MSTR's current bare pneumatic tube (BPT) and cadmium lined pneumatic tube (CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g, irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals who do not have on-site access to a nuclear reactor facility to remotely participate in research and educational activities.

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

OpenAIRE

Romero Sayago, Sara Isabel

2016-01-01

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

The U.S. reduced enrichment research and test reactor (RERTR) program

International Nuclear Information System (INIS)

Travelli, A.

1993-01-01

Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

SPEED COMPLIANCE IN FREEWAY VARIABLE SPEED LIMIT SYSTEM – CASE STUDY OF THE PRAGUE CITY RING

Directory of Open Access Journals (Sweden)

Michał MATOWICKI

2016-03-01

Full Text Available Many previous studies have confirmed the strong relationship between speed compliance and the frequency and severity of traffic accidents. Variable speed limit (VSL system as a measure to improve traffic safety enables the freeway system to change its posted speed limit based on various traffic and environmental conditions. Such system helps drivers to recognize the upcoming events, to adjust their driving style and in such way to address speed variation of the traffic flow. This is called speed harmonization. Although many studies researching the effect of VSL system on the traffic stream can be found, there are only few addressing its influence on the drivers behavior, particularly focusing on their tolerance limit and compliance, which has crucial meaning for future design of controlling algorithms. This study was prepared to inspect this grey area by studying the data from the VSL system at Prague city ring, describing the influence of the highway management system and its influence on drivers.

Status of international environmental remediation activities: A report from the Prague conference

International Nuclear Information System (INIS)

Slate, S.C.; Thornhill, C.K.; Allen, R.E.

1993-10-01

The Prague Conference on nuclear waste management and environmental remediation provided extensive interchange of ideas and insight into new technologies and management approaches throughout the world. A variety of environmental remediation technologies have potential application to Department of Energy facilities; others illustrate pitfalls to be avoided. This paper presents the highlights from the first environmental remediation (ER) technical program in the American Society of Mechanical Engineers' series of international nuclear waste management conferences. This program covers ER technologies, decontamination and decommissioning (D ampersand D) technologies and experience, ER site characterization and modeling, management of and results from actual clean up actions, and data on several major international environmental problems. Focusing on direct benefits to the Department of Energy's (DOE) ER Program, this paper summarizes pertinent technical information, identifies useful technical papers, lists key technical contacts, and identifies specific actions to obtain additional information. US attendance at meetings like this is normally quite limited compared to attendance at North American meetings. The purpose of this paper then is to increase general awareness of this meeting in US technical circles and to broadly disseminate key information to US ER programs and contractors. To do this, the paper is organized to present background information on the conference itself, document the beneficial technical information, and outline ongoing information exchange activities

NATO's new strategic concept and the future of nuclear disarmament in Europe; Das Neue Strategische Konzept der NATO und die Zukunft der nuklearen Abruestung in Europa

Energy Technology Data Exchange (ETDEWEB)

Dembinski, Matthias; Mueller, Harald

2010-07-01

The study on NATO's new strategic concept and the future of nuclear disarmament in Europe covers the following topics: From the Prague speech to the strategic concept - the significance and role of the substrategic nuclear weapons in Europe: the substrategic nuclear weapons (SSNW) in the USA: relict of the Cold war or clamps of the transatlantic security? The future of the SSNW from the Prague speech to the strategic concept. The positions of selected NATO members and Russia. Conclusions and recommendations: From the significance of the US nuclear weapons in Europe. Arms control policy options.

Reactor container

International Nuclear Information System (INIS)

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

1997-01-01

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

Activity on non-destructive testing as constituent element of the quality management in accordance with ISO 9001:2000 standard at The Institute of Nuclear Physics, Kazakhstan

International Nuclear Information System (INIS)

Kadyrzhanov, K.K.; Kislitsin, S.B.; Ablanov, M.B.

2004-01-01

An increase of technical and public safety requirements for facilities of nuclear industries, an efficient quality control based on non-destructive testing (NDT) techniques is crucial. Therefore, Institute of Nuclear Physics (INP) through NDT Division makes efforts towards a competent NDT inspection of its facilities starting from research reactor of WWR-K type with a further activity according to the National Program for Development in Nuclear Industry. The additional objective is to harmonize the present codes and standards for Nuclear Industry as an integral part of the INP policy in a quality management according ISO 9001:2000 Standard. (author)

Smoking Habits and Attitudes in Students of the Third Faculty of Medicine of Charles University in Prague.

Science.gov (United States)

Schneidrová, Dagmar; Herotová, Tereza Kopřivová; Šustková, Magdaléna; Hynčica, Viktor

2016-06-01

This study seeks to assess smoking habits, attitudes and intention to quit in students of the Third Faculty of Medicine of Charles University in Prague, Czech Republic. A cross-sectional survey designed to obtain information on smoking history, current smoking status, cessation attempts, and attitudes towards smoking among health professionals was conducted in 452 students of the first and last years of a 6-year Master's Study Programme (General Medicine) and a 3-year Bachelor's Study Programme (Public Health). An anonymous questionnaire was administered during the classes in the course of academic years 2011-12 and 2012-13. 5.7% of the Master's Study Programme students (3.3% women and 9.0% men ) and 4.8% of the Bachelor's Study Programme students reported that they are regular smokers. The share of regular smokers was almost twice as big in students of the English Curriculum of the Master's Programme (10.7%) in comparison with the students of the Czech Curriculum (4.5%), and more than twice as big in students of the last years of both study programmes (3.9% in students of the 1st year and 10.8% in students of the 6th year of the Master's Programme; 3.2% in students of the 1st year and 7.0% in students of the 3rd year of the Bachelor's Study Programme). At the time of the research, 18.9% of students of the Master's Programme and 17.1% of students of the Bachelor's Programme were occasional smokers. 5.9% of students of the Master's Programme and 19.0% of students of the Bachelor's Programme reported that they quit smoking during their studies at the medical faculty; on the contrary, 9.8% of students of the Master's Programme and 14.3% of students of the Bachelor's Programme started smoking during that time. Smoking in health professionals undermines their significant role in health promotion and prevention of chronic diseases in their patients. Therefore, education at the medical faculty should focus on motivation of future health professionals towards non

U.S. progress in the development of very high density low enrichment research reactor fuels

International Nuclear Information System (INIS)

Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

2012-01-01

The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

Tank type reactor

International Nuclear Information System (INIS)

Otsuka, Fumio.

1989-01-01

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

Advanced reactor development: The LMR integral fast reactor program at Argonne

International Nuclear Information System (INIS)

Till, C.E.

1990-01-01

Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

A Compilation of Boiling Water Reactor Operational Experience for the United Kingdom's Office for Nuclear Regulation's Advanced Boiling Water Reactor Generic Design Assessment

Energy Technology Data Exchange (ETDEWEB)

Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Liao, Huafei [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-12-01

United States nuclear power plant Licensee Event Reports (LERs), submitted to the United States Nuclear Regulatory Commission (NRC) under law as required by 10 CFR 50.72 and 50.73 were evaluated for reliance to the United Kingdom’s Health and Safety Executive – Office for Nuclear Regulation’s (ONR) general design assessment of the Advanced Boiling Water Reactor (ABWR) design. An NRC compendium of LERs, compiled by Idaho National Laboratory over the time period January 1, 2000 through March 31, 2014, were sorted by BWR safety system and sorted into two categories: those events leading to a SCRAM, and those events which constituted a safety system failure. The LERs were then evaluated as to the relevance of the operational experience to the ABWR design.

Inspection program for U.S. research reactors

International Nuclear Information System (INIS)

Isaac, Patrick J.

2010-01-01

This paper presents an established program for inspection of nuclear research reactors to ensure that systems and techniques are in accordance with regulatory requirements and to provide protection for the health and safety of the public. The inspection program, implemented from the time a facility gets licensed, remains in effect through operations, shutdown, decommissioning, and until the license is terminated. The program establishes inspection methodology for operating, safeguards, and decommissioning activities. Using a performance- based approach, inspectors focus their attention on activities important to safety. Inspection procedures allow the inspectors to assess facility safety and compliance to applicable requirements. A well designed inspection program is an integral part of the mechanism to ensure that the level of performance in the strategic areas of reactor safety, radiation safety, and safeguards is acceptable and provides adequate protection of public health and safety. (author)

Centralizace vědy – státní společenskovědní ústavy a vznik ČSAV. Osudy Slovanského a Orientálního ústavu v letech 1945–1953

Czech Academy of Sciences Publication Activity Database

Jůnová Macková, Adéla

2015-01-01

Roč. 7, č. 2 (2015), s. 167-192 ISSN 1803-9448 Institutional support: RVO:67985921 Keywords : Slavonic Institute, Prague * Oriental Institute, Prague * Czechoslovak Academy of Science s Subject RIV: AB - History

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Planning the Decommissioning of Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

2013-08-15

In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

Science.gov (United States)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

Design, fabrication and transportation of Si rotating device

International Nuclear Information System (INIS)

Kimura, Nobuaki; Imaizumi, Tomomi; Takemoto, Noriyuki; Tanimoto, Masataka; Saito, Takashi; Hori, Naohiko; Tsuchiya, Kunihiko; Romanova, Nataliya; Gizatulin, Shamil; Martyushov, Alexandr; Nakipov, Darkhan; Chakrov, Petr; Tanaka, Futoshi; Nakajima, Takeshi

2012-06-01

Si semiconductor production by Neutron Transmutation Doping (NTD) method using the Japan Materials Testing Reactor (JMTR) has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency (JAEA) in order to expand industry use. As a part of investigations, irradiation test of silicon ingot for development of NTD-Si with high quality was planned using WWR-K in Institute of Nuclear Physics (INP), National Nuclear Center of Republic of Kazakhstan (NNC-RK) based on one of specific topics of cooperation (STC), Irradiation Technology for NTD-Si (STC No.II-4), on the implementing arrangement between NNC-RK and the JAEA for 'Nuclear Technology on Testing/Research Reactors' in cooperation in research and development in nuclear energy and technology. As for the irradiation test, Si rotating device was fabricated in JAEA, and the fabricated device was transported with irradiation specimens from JAEA to INP-NNC-RK. This report described the design, the fabrication, the performance test of the Si rotating device and transportation procedures. (author)

Revised reactor accident source terms in the U.S. and implementation for light water reactors

International Nuclear Information System (INIS)

Soffer, L.; Lee, J.Y.

1992-01-01

Current NRC reactor accident source terms used for licensing are contained in Regulatory Guides 1.3 and 1.4 and specify that 100 % of the core inventory of noble gases and 25 % of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental (I 2 ) iodine. These assumptions have strongly affected present nuclear plant designs. Severe accident research results have confirmed that although the current source term is very substantial and has resulted in a very high level of plant capability, the present source term is no longer compatible with a realistic understanding of severe accidents. The NRC has issued a proposed revision of the reactor accident source terms as part of several regulatory activities to incorporate severe accident insights for future plants. A revision to 10 CFR 100 is also being proposed to specify site criteria directly and to eliminate source terms and doses for site evaluation. Reactor source terms will continue to be important in evaluating plant designs. Although intended primarily for future plants, existing and evolutionary power plants may voluntarily apply revised accident source term insights as well in licensing. The proposed revised accident source terms are presented in terms of fission product composition, magnitude, timing and iodine chemical form. Some implications for light water reactors are discussed. (author)

An overview of the U.S. Department of Energy Experimental Boiling Water Reactor Decontamination and Decommissioning Project

International Nuclear Information System (INIS)

Murphie, W.E.; Mckernan, M.L.

1991-01-01

This paper provides an overview of the U.S. Department of Energy's (DOE) Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D and D) Project. Physical decommissioning work started in 1986 and is scheduled for completion in 1994. The project total estimated cost is 14.3 million (1990, U.S.) dollars. The reactor pressure vessel will be removed by segmentation. Another notable project feature is that D and D operations were planned for and carried out with a small work force comprised of four to six D and D laborers, one or two health physics technicians, an engineer, and a project manager. When the D and D work is completed the facility will be recycled for other productive uses. (author)

Use of enriched uranium in Canada's power reactors

International Nuclear Information System (INIS)

Dormuth, K.W.; Jackson, D.P.

2011-01-01

Recent trends in Canadian nuclear power reactor design and proposed development of nuclear power in Canada have indicated the possibility that Canada will break with its tradition of natural uranium fuelled systems, designed for superior neutron economy and, hence, superior uranium utilization. For instance, the Darlington B new reactor project procurement process included three reactor designs, all employing enriched fuel, although a natural uranium reactor design was included at a late stage in the ensuing environmental assessment for the project as an alternative technology. An evaluation of the alternative designs should include an assessment of the environmental implications through the entire fuel cycle, which unfortunately is not required by the environmental assessment process. Examples of comparative environmental implications of the reactor designs throughout the fuel cycle indicate the importance of these considerations when making a design selection. As Canada does not have enrichment capability, a move toward the use of enriched fuel would mean that Canada would be exporting natural uranium and buying back enriched uranium with value added. From a waste management perspective, Canada would need to deal with mill, refinery, and conversion tailings, as well as with the used fuel from its own reactors, while the enrichment supplier would retain depleted uranium with some commercial value. On the basis of reasoned estimates based on publicly available information, it is expected that enrichment in Canada is likely to be more profitable than exporting natural uranium and buying back enriched uranium. Further, on the basis of environmental assessments for enrichment facilities in other countries, it is expected that an environmental assessment of a properly sited enrichment facility would result in approval. (author)

Advanced CANDU reactors

International Nuclear Information System (INIS)

Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

1988-12-01

AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

K-East and K-West Reactors

Data.gov (United States)

Federal Laboratory Consortium — Hanford's "sister reactors", the K-East and the K-West Reactors, were built side-by-side in the early 1950's. The two reactors went operational within four months of...

René Wellek a Pražská škola

Czech Academy of Sciences Publication Activity Database

Sládek, Ondřej

2016-01-01

Roč. 1, č. 1 (2016), s. 19-30 ISSN 2453-8507 Institutional support: RVO:68378068 Keywords : literary theory ; structuralism ; poetics ; Wellek, René ; Prague School * structuralism * poetics * Wellek, René * Prague School * Prague linguistic circle * Mukařovský, Jan Subject RIV: AJ - Letters, Mass-media, Audiovision

Reactor pressure tank

International Nuclear Information System (INIS)

Dorner, H.; Scholz, M.; Jungmann, A.

1975-01-01

In a reactor pressure tank for a nuclear reactor, self-locking hooks engage a steel ring disposed over the removable cover of the steel vessel. The hooks exert force upon the cover to maintain the cover in a closed position during operation of the reactor pressure tank. The force upon the removal cover is partly the result of the increasing temperature and thermal expansion of the steel vessel during operation. The steel vessel is surrounded by a reinforced-concrete tank. (U.S.)

Education of 'nuclear' students (BSc and MSc curricula) at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague

International Nuclear Information System (INIS)

Matejka, K.; Zeman, J.

2003-01-01

The Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague has been educating nuclear power engineering experts for nearly half a century. The article describes the current status and prospects of education of new specialists at the faculty for all nuclear power-related areas within the MSc and BSc level curricula. The current transition to 'European type' structured education, enabling students who have graduated from the BSc programme to continue smoothly their MSc programme, is outlined. The major courses of the 'Nuclear Engineering' educational specialisation, focused on nuclear power, environment, and dosimetry, are highlighted, including the number of lessons taught in each study year. (author)

Application of S-CO{sub 2} Cycle for Small Modular Reactor coupled with Desalination System

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-10-15

The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO{sub 2} power cycle technology. The S-CO{sub 2} Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO{sub 2} cycles for SMART with desalination system is conducted. The simple recuperated S-CO{sub 2} cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Luncheon address: Development of the CANDU reactor

International Nuclear Information System (INIS)

Bain, A.S.

1997-01-01

The paper is a highlight of the some of the achievements in the development of the CANDU Reactor, taken from the book C anada Enters the Nuclear Age . The CANDU reactor is one of Canada's greatest scientific/engineering achievements, that started in the 1940's and bore fruit with the reactors of the 60's, 70's, and 80's. The Government decided in the 1950's to proceed with a demonstration nuclear power reactor (NPD), AECL invited 7 Canadian corporations to bid on a contract to design and construct the NPD plant. General Electric was selected. A utility was also essential for participation and Ontario Hydro was chosen. In May 1957 it was concluded that the minimum commercial size would be about 200MWe and it should use horizontal pressure tubes to contain the fuel and pressurized heavy water coolant. The book also talks of standard out-reactor components such as pumps, valves, steam generators and piping. A major in-reactor component of interest was the fuel, fuel channels and pressure tubes. A very high level of cooperation was required for the success of the CANDU program

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

German research reactor back-end provisions

International Nuclear Information System (INIS)

Koester, Siegfried; Gruber, Gerhard

2002-01-01

Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

Research reactor safety - an overview of crucial aspects

Energy Technology Data Exchange (ETDEWEB)

Laverie, M. [Atomic Energy Commission, Saclay, F-91191 Gif sur Yvette (France)

1998-07-01

Chronology of the commissioning orders of the French research reactors illustrates the importance of the time factor. When looking at older reactors, one must, on one hand, demonstrate, not only the absence of risks tied to the reactor's ageing, but, on the other hand, adapt the reactor's original technical designs to today's safety practices and standards. The evolution of reactor safety requirements over the last twenty years sometimes makes this adaptation difficult. The design of the next research reactors, after a one to two decades pause in construction, will require to set up new safety assessment bases that will have to take into account the nuclear power plant safety evolution. As a general statement, research reactor safety approaches will require the incorporation of specific design rules for research reactors: experience feedback for one of a kind design, frequent modifications required by research programmes, special operational requirements with operators/researchers interfaces. (author)

Meeting the reactor operator's information needs using functional analysis

Energy Technology Data Exchange (ETDEWEB)

Nelson, W.R.; Clark, M.T.

1980-01-01

Since the accident at Three Mile Island, many ideas have been proposed for assisting the reactor operator during emergency situations. However, some of the suggested remedies do not alleviate an important shortcoming of the TMI control room: the operators were not presented with the information they needed in a manner which would allow prompt diagnosis of the problem. To address this problem, functional analysis is being applied at the LOFT facility to ensure that the operator's information needs are being met in his procedures and graphic displays. This paper summarizes the current applications of functional analysis at LOFT.

Analysis of application of alternative drive systems for international heavy-duty transport on Wroclaw-Dresden-Prague routes

Science.gov (United States)

Skrętowicz, Maria; Sroka, Zbigniew

2017-11-01

The depletion of the fossil fuels resources, significant increase of the air pollution caused by the use of internal combustion engines, and emission of carbon dioxide which is responsible for the greenhouse effect escalates the development of vehicle's alternative drive systems. Generally, the emphasis is given to the alternative fuels (natural gas CNG, mixture of propane-butane gases LPG, hydrogen, alcohol fuels, biofuels) and hybrid or electric vehicles. Roads between large industrial and commercial centres, i.e. Wroclaw - Dresden - Prague, are used mainly by heavy-duty vehicles. Consequently, the contribution of the road transport to the ecological threat in this realm is significant. The objectives of this research were the assessment of the traffic volume and emission rate of exhaust gases caused by heavy-duty vehicles on the analysed roads and evaluation of the possibility of using existing and alternative drive systems in vehicles driving on the roads in the analysed region.

The nuclear question at the start of the '80s: the breeder reactor

International Nuclear Information System (INIS)

Owen, R.; Svensson, B.

1980-01-01

The four newspaper articles and the letter cover the following matters: general introduction about breeder reactors and the situation in Swedish politics; visit to Dounreay to discuss breeder reactors (breeding, safety, plutonium production, radiation protection); PuO 2 -UO 2 mixed fuel; description of breeder reactors; efficiency in use of U-235; DFR and PFR; breeder reactors in Swedish politics (arguments for and against nuclear power in general, breeder reactors in particular); discussion of the future of nuclear power in Sweden. (U.K.)

78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

Science.gov (United States)

2013-11-29

... NUCLEAR REGULATORY COMMISSION [NRC-2013-0260] Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this Federal Register notice to inform the public of a slight change in the manner of distribution of publicly available operating reactor licensing...

Design and construction of a preamplifier for research reactor control system using Russia’s neutron detectors

International Nuclear Information System (INIS)

Trinh Dinh Hai; Vo Van Tai; Le Van Diep; Nguyen Nhi Dien

2016-01-01

This paper presents the design and construction of a preamplifier device for Research Reactor Control System, using Russia’s Neutron Detectors of ionization and fission chambers. In this work, the preamplifier device which consists of a wide range Current to Frequency Converter block used with a compensation ionization chamber type KNK-3 to measure the thermal neutron flux in the range of 1x10"6 - 1x10"1"1 n/cm"2.s, a Pulse Preamplifier block used with a fission chamber type KNK-15 to measure the thermal neutron flux in the range of 1x10"0 - 1x10"6 n/cm"2.s, and a Power Supply block, was designed and tested in different conditions in the laboratory and at Dalat Nuclear Research Reactor (DNRR). Obtained results show that, the above blocks have almost design specifications as equivalent or better in comparison with the same function blocks of the DNRR Control System which were designed by the former Soviet Union. They also meet the utilization requirements as well as the experimental and training purposes. (author)

Diamond as a solid state micro-fission chamber for thermal neutron detection at the VR-1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Pomorski, Michal; Mer-Calfati, Christine [CEA-LIST, Diamond Sensors Laboratory, 91191, Gif-sur-Yvette (France); Foulon, Francois [CEA, National Institute for Nuclear Science and Technology, 91191, Gif-sur-Yvette (France); Sklenka, Lubomir; Rataj, Jan; Bily, Tomas [Department of Nuclear Reactors,Faculty of Nuclear Science and Physical Engineering, Czech Technical University, V. Holesovickach 2, 180 00 PRAHA 8 (Czech Republic)

2015-07-01

Prague. The Training Reactor VR-1 is a pool type (light water) reactor based on UO{sub 2} low enriched uranium. It has a nominal power of 1 kW, and can be operated for a short period up to 5 kW. The arrangement of the reactor pool reactor facilitates access to the core, setting and removal of various experimental samples and detectors, and safe and easy handling of fuel assemblies. The reactor is equipped with two horizontal channels (radial and tangential) and 10 vertical channels, of varying diameters, which can be loaded into various core positions, and one pneumatic transfer system. It is also equipped with several specifically designed educational instrumentation systems that can be used to supply complementary measurements and characterization around the reactor. The reactor is operated by the Department of Nuclear Reactors of the Faculty of Nuclear Sciences and Physical Engineering of the Czech Technical University in Prague. The two detectors were placed in-core through one of the vertical insertion channel. They were coupled to remote placed (5 m BNC cable) classical nuclear charge sensitive electronics. Detection properties of both sensors, including: pulse height spectra of U-235 fission fragments (response linearity with neutron flux, count rate, gamma background, were evaluated varying the power of the reactor from 0.005 W to 500 W. The evolution of the counting rate of the thinned optical grade detector as a function of counting rate of a gas ionization chamber used currently for reactor monitoring shows the very good linearity of the detector over the 5 decades. Similar results were obtained with the PIM detector. Additionally fast transient current signals of the detectors were recorded on a digital storage oscilloscope (DSO) using broad-band amplifier and with a simple bias-T, showing potential use of such sensors for neutron counting with no need of an amplification stage, since non-amplified signals from fission fragments exceeded 4 mV in amplitude

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

Nuclear Power: Outlook for New U.S. Reactors

National Research Council Canada - National Science Library

Parker, Larry; Holt, Mark

2007-01-01

.... The renewed interest in nuclear power has resulted primarily from higher prices for natural gas, improved operation of existing reactors, and uncertainty about future restrictions on coal emissions...

Reactor power measuring device

International Nuclear Information System (INIS)

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

1993-01-01

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

Reactor technology

International Nuclear Information System (INIS)

Erdoes, P.

1977-01-01

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

Advanced Reactor Development in the United States

Energy Technology Data Exchange (ETDEWEB)

Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

1990-04-15

In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

Energy Technology Data Exchange (ETDEWEB)

Haley, T.C. [Rensselaer Polytechnic Inst., Troy, NY (United States)

2011-07-01

Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

IAEA Mission Sees High Commitment to Safety at Ghana's Research Reactor After HEU to LEU Fuel Conversion

International Nuclear Information System (INIS)

2018-01-01

An International Atomic Energy Agency (IAEA) team of experts said the operator of Ghana’s research reactor has demonstrated a high commitment to safety following the conversion of the reactor core to use low enriched uranium (LEU) as fuel instead of high enriched uranium (HEU). The team also made recommendations for further safety enhancements. The Integrated Safety Assessment for Research Reactors (INSARR) team concluded a five-day mission today to assess the safety of the GHARR-1 research reactor, originally commissioned in 1994. The 30 kW reactor, operated by the Ghana Atomic Energy Commission (GAEC) at the National Nuclear Research Institute in the capital Accra, is used primarily for trace element analysis for industrial or agricultural purposes, research, education and training. In 2017, the reactor core was converted in a joint effort by Ghana, the United States and China, with assistance from the IAEA. The IAEA supported the operation to eliminate proliferation risks associated with HEU, while maintaining important scientific research. The team made recommendations for improvements to the GAEC, including: • Completing the revision of reactor safety and operating documents to reflect the results of the commissioning of the reactor after the core fuel conversion. • Enhancing the training and qualification programme for operating personnel. • Improving the capability for monitoring operational safety parameters under all conditions. • Strengthening radiation protection by establishing an effective radiation monitoring of workplace. The GAEC said it will request a follow-up INSARR mission by 2020.

Solid State Reactor Final Report

Energy Technology Data Exchange (ETDEWEB)

Mays, G.T.

2004-03-10

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

Super critical water reactors

International Nuclear Information System (INIS)

Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

2005-01-01

Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2014-01-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

2014-07-01

Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

Quantitative allochem compositional analysis of Lochkovian-Pragian boundary sections in the Prague Basin (Czech Republic)

Science.gov (United States)

Weinerová, Hedvika; Hron, Karel; Bábek, Ondřej; Šimíček, Daniel; Hladil, Jindřich

2017-06-01

Quantitative allochem compositional trends across the Lochkovian-Pragian boundary Event were examined at three sections recording the proximal to more distal carbonate ramp environment of the Prague Basin. Multivariate statistical methods (principal component analysis, correspondence analysis, cluster analysis) of point-counted thin section data were used to reconstruct facies stacking patterns and sea-level history. Both the closed-nature allochem percentages and their centred log-ratio (clr) coordinates were used. Both these approaches allow for distinguishing of lowstand, transgressive and highstand system tracts within the Praha Formation, which show gradual transition from crinoid-dominated facies deposited above the storm wave base to dacryoconarid-dominated facies of deep-water environment below the storm wave base. Quantitative compositional data also indicate progradative-retrogradative trends in the macrolithologically monotonous shallow-water succession and enable its stratigraphic correlation with successions from deeper-water environments. Generally, the stratigraphic trends of the clr data are more sensitive to subtle changes in allochem composition in comparison to the results based on raw data. A heterozoan-dominated allochem association in shallow-water environments of the Praha Formation supports the carbonate ramp environment assumed by previous authors.

Plant maintenance and advanced reactors, 2006

Energy Technology Data Exchange (ETDEWEB)

Agnihotri, Newal (ed.)

2006-09-15

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

U.S. activities related to fast reactors and ADS

International Nuclear Information System (INIS)

Finck, Phillip J.

2001-01-01

The U.S. nuclear power industry achieved its second straight year of record power generation levels during 2000. Total power generated was 753.9 billion kWh, 3.5 percent above the previous record of 728.1 billion kWh set in 1999. Recent years of reliable service and declining production costs have led to increased industry interest in nuclear power plant license renewal. The Nuclear Regulatory Commission approved the license renewal applications for five U.S. nuclear power plants in 2000. Five additional applications were received and 28 more are planned to be submitted by 2004. Concerns over energy resource availability, climate change, air quality, and energy security suggest an important role for nuclear power in future energy supplies. While the current Generation II and III nuclear power plant designs provide an economically, technically, and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. To explore these opportunities, the U. S. Department of Energy's Office of Nuclear Energy, Science and Technology has engaged governments, industry, and the research community worldwide in a wide-ranging discussion on the development of next-generation nuclear energy systems known as 'Generation IV'. The I-NERI program has the following objectives: Develop advanced concepts and scientific breakthroughs in nuclear fission and reactor technology to address and overcome the principal technical and scientific obstacles to the expanded use of nuclear energy worldwide; Promote bilateral and multilateral collaboration with international agencies and research organizations to improve development of nuclear energy; and Promote and maintain nuclear science and engineering infrastructure to meet future technical challenges. I-NERI will sponsor innovative research and development in the following areas: Next-generation nuclear energy plant designs with higher efficiency

Reactor power control device

International Nuclear Information System (INIS)

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

1998-01-01

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

The Simulator Development for RDE Reactor

Science.gov (United States)

Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

Research reactor job analysis - A project description

International Nuclear Information System (INIS)

Yoder, John; Bessler, Nancy J.

1988-01-01

Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

The Use of a Consumer Survey to Determine the Relevant Market – Case Study for public transport between Prague and Most

Directory of Open Access Journals (Sweden)

Kleinová Eliška

2016-03-01

Full Text Available The aim of this paper is to define a relevant market in the passenger transportation industry for the route between Prague and Most. A rising number of bus companies operating on this particular market suggests competition pressures. On the basis of European Commission legislation and its common practice we apply a demand-side substitution analysis in the form of the so called SSNIP test. Data for the empirical analysis were collected by means of a customer survey, which captured reactions to a 10% increase in the price of purchased tickets. The survey outcomes were then used to calculate an own price elasticity of demand and to carry out a critical loss analysis in order to define the relevant market.

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

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

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Nuclear reactors built, being built, or planned 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

Nuclear reactors built, being built, or planned: 1996

International Nuclear Information System (INIS)

1997-08-01

This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled

Challenges in licensing a sodium-cooled advanced recycling reactor

International Nuclear Information System (INIS)

Levin, Alan E.

2008-01-01

As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

MLR reactor

International Nuclear Information System (INIS)

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

1998-01-01

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

On the yield of cold and ultracold neutrons for liquid hydrogen at low temperatures near the melting point

CERN Document Server

Morishima, N

1999-01-01

The neutron scattering cross sections for liquid hydrogen in the temperature range from the melting point to the boiling point are calculated. It is shown that lowering the temperature results in a significant increase in the yield of cold neutrons: for instance, a 44% increase for an incident neutron energy of 19.4 meV. The major cause of this increment is the para-to-ortho transition of a hydrogen molecule though accompanied by an appreciable increase in the density. The results of the cold- and ultracold-neutron yields are discussed in connection with the experimental results of Altarev et al. at the WWR-M reactor.

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.

1976-01-01

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

Nuclear Power: Outlook for New U.S. Reactors

National Research Council Canada - National Science Library

Parker, Larry; Holt, Mark

2007-01-01

Nearly three decades after the most recent order was placed for a new nuclear power plant in the United States, several utilities are now expressing interest in building a total of up to 30 new reactors...

''Sleeping reactor'' irradiations: Shutdown reactor determination of short-lived activation products

International Nuclear Information System (INIS)

Jerde, E.A.; Glasgow, D.C.

1998-01-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux (φ) of ∼ 4 x 10 14 n/cm 2 · s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of ∼ 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about ± 0.5 s) make irradiations of 9 Be(γ,n) 8 Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to ∼ 1 x 10 10 n/cm 2 · s within 1 h. By the time the fuel elements are removed, the flux has dropped to ∼ 6 x 10 8 . Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant

U.S. fast reactor materials and structures program

International Nuclear Information System (INIS)

Harms, W.O.; Purdy, C.M.

1984-01-01

The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

International Nuclear Information System (INIS)

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

1997-03-01

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here

Significance assessment of small-medium sized reactors

Energy Technology Data Exchange (ETDEWEB)

Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

2002-12-01

Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

Proceedings of the CNRA Workshop on New Reactor Siting, Licensing and Construction Experience

International Nuclear Information System (INIS)

2011-01-01

This report documents the proceedings from the 'Workshop on New Reactor Siting, Licensing and Construction Experience', held in Prague, Czech Republic on 15-17 September 2010. A total of 59 specialists from 16 countries and international organisations attended. The Meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the State Office for Nuclear Safety (SUJB) in Czech Republic. The objectives of the workshop were to review and discuss recent and past construction experience lessons learned including perspectives from regulatory authorities, as well as vendors, and licensee. The workshop addressed issues associated with project management resources including: a) overall human resources, expertise, experience and organisation available to the licensee, b) capability of each potential vendor (in-house knowledge and skills versus planned subcontracting and subcontractor management). The workshop also discussed the lessons learned in the regulation of site selection, evaluation and site preparation as well as the review of regulatory practices for the licensing of new reactors, including the regulatory body infrastructure, staffing and expertise needed. The workshop provided an excellent opportunity to communicate recent experience on these topics to a wider audience, including participants from OECD member countries as well as New Entrants from non-OECD member countries. The workshop allowed the WGRNR group to introduce and discuss the current programme of work and products under development in order to gain insights from workshop participants on each of the programme of work areas, and get feedback on additional focus areas. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to presentations of the licensing process for new reactors followed by different member countries. The second technical session was

Lessons learned from 50 years period the storage of the spent fuel from nuclear research reactor VVR-S

International Nuclear Information System (INIS)

Dragusin, M.

2010-01-01

The nuclear research reactor VVR-S was commissioned in July 1957. This reactor is in permanent shutdown since December 1997 and will be decommissioned. The duration of the decommissioning project is 11 years. The first year of decommissioning project is 2010. The spent nuclear fuels resulting from the 40 years of operating the nuclear research reactor are stored under wet conditions. The chemical and physical water parameters monitored are: transparency, conductibility, pH, chloride content, oxygen content, temperature, dry residual content, Al, Mn, Mg, Fe, Vn, Cr. Residual dry content must be maintained in requested range in order to prevent degradation and corrosion both of the clads, assemblies and linen material of the ponds. Two types of the nuclear fuel assemblies were used: LEU type -EK-10 and HEU type S-36 Russian origin. All spent nuclear fuel assemblies HEU-S-36 type were repatriated in Russian Federation in June 2009 in safety and security conditions without any problems due of the wet storage, after 25 years storage in wet conditions. The spent nuclear fuel assemblies types LEU EK-10 were stored in wet conditions more than 50 years. This paper describes the lessons learned during the 50 years management of the spent nuclear fuel resulted from the operation the research reactor VVR-S. The management was based on the maintenance of water parameters by water filtration, using at all times air HEPA filter incorporated in technological ventilation system and by monitoring the level, temperature, physical and chemical parameters of the water storage from ponds and by controlling ponds linen physical integrity. Also we have used the discs having the same compositions with materials from assemblies stored in the same ponds, in order to verify degradation and corrosion phenomena induced due to the quality of storage water. The paper will described these results obtained by metallographic, visual, XRF analysis onto discs and dry residual samples from storage

[History of the 4th Department of Internal Medicine of the First Faculty of Medicine at Charles University and the General University Hospital in Prague].

Science.gov (United States)

Bartůněk, Petr

In 2015, the doctors and nurses of the 4th Department of Internal Medicine of the First Faculty of Medicine, Charles University and the General University Hospital in Prague celebrated the 70th anniversary of its founding. The article summarizes the clinics contribution to the field of internal medicine, and particularly to angiology, hepatogastroenterology and lipidology. It comments the clinics current activities and the possibilities of its further development. Attention is also paid to the tradition of high ethical and professional standards of medical care in accordance with the norms established by the clinic's founder, prof. MUDr. Bohumil Prusík.

Strategy to support HTGR fuel for the 10 MW Indonesia’s experimental power reactor (RDE)

International Nuclear Information System (INIS)

Taswanda Taryo; Geni Rina Sunaryo; Ridwan; Meniek Rachmawati

2018-01-01

The Indonesia’s 10 MW experimental power reactor (RDE) is developed based on high temperature gas-cooled reactor (HTGR) and the program of the RDE was firstly introduced to the Agency for National Development Planning (BAPPENAS) at the beginning of 2014. The RDE program is expected to have positive impacts on community prosperity, self-reliance and sovereignty of Indonesia. The availability of RDE will be able to accelerate advanced nuclear power technology development and hence elevate Indonesia to be the nuclear champion in the ASEAN region. The RDE is expected to be operable in 2022/2023. In terms of fuel supply for the reactor, the first batch of RDE fuel will be inclusive in the RDE engineering, procurement and construction (RDE-EPC) contract for the assurance of the RDE reactor operation from 2023 to 2027. Consideration of RDE fuel plant construction is important as RDE can be the basis for the development of reactors of similar type with small-medium power(25 MWe–200/300 MWe), which are preferable for eastern part of Indonesia. To study the feasibility of the construction of RDE fuel plant, current state of the art of the R&D on HTGR fuel in some advanced countries such as European countries, the United States, South Africa and Japan will be discussed and overviewed to draw a conclusion about the prospective countries for supporting the fuel for long-term RDE operation. The strategy and road map for the preparation of the RDE fuel plant construction with the involvement of national stake holders have been developed. The best possible vendor country to support HTGR fuel for long-term operation is finally accomplished. In the end, this paper can be assigned as a reference for the planning and construction of HTGR RDE fuel fabrication plant in Indonesia. (author)

Perception of neighborhood environment and health risk behaviors in Prague's teenagers: a pilot study in a post-communist city.

Science.gov (United States)

Spilkova, Jana; Dzúrova, Dagmar; Pitonak, Michal

2014-10-14

A youths' neighborhood can play an important role in their physical, health, and emotional development. The prevalence of health risk behavior (HRB) in Czech youth such as smoking, drug and alcohol use is the highest in Europe. To analyze differences in HRB in youth residents within different types of Prague's neighborhoods in relation to the perception of the built environment, quality of their school and home environments. The data is based on the on-line survey among elementary school students aged between 14-15 years, which was administered in19 selected schools in Prague, during the months of October 2013 to March 2014. Respondents were asked their opinions on various issues related to their HRB, about their indoor and outdoor housing and school environments. The questionnaire was completed by 407 students. Factor analysis with a principal components extraction was applied to determine the underlying structure in the variables. A consequent field research was conducted to map the opportunity hot spots and critical places around the elementary schools. Binge drinking has been reported mainly by the students living in the housing estates with blocks of flats. The most frequent occurrence of daily smokers was found in the neighborhoods of old city apartment houses. High prevalence of risky marijuana use almost in all the surveyed types of neighborhoods. The respondents were more critical in their evaluation of school characteristics. The neighborhoods critically evaluated by the students as regards the school outdoor environments were the older apartment houses in the historical centre and inner city, the school indoor environment was worst assessed within the housing estate neighborhoods. Our results suggest that perceptions of problems in both residential and school environment are associated with HRB. This fact makes this issue of a serious importance also from the policy point of view. Mainly the school surroundings have to be better managed by the local

Status of the advanced boiling water reactor and simplified boiling water reactor

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

Research reactors: a tool for science and medicine; Reactores de investigacion: herramientas para la ciencia y la medicina

Energy Technology Data Exchange (ETDEWEB)

Ordonez, Juan [Investigacion Aplicada SE (INVAP), San Carlos de Bariloche (Argentina)

2001-07-01

The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

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)

LBB application in the US operating and advanced reactors

Energy Technology Data Exchange (ETDEWEB)

Wichman, K.; Tsao, J.; Mayfield, M.

1997-04-01

The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

U.S. Department of Energy University Reactor Sharing Program at the University of Florida. Final report for period August 15, 2000 - May 31, 2001

Energy Technology Data Exchange (ETDEWEB)

Vernetson, William G.

2002-01-01

Department of Energy Grant Number DE-FG02-96NE38152 was supplied to the University of Florida Training Reactor (UFTR) facility through the U.S. Department of Energy's University Reactor Sharing Program. The renewal proposal submitted in January 2000 originally requested over $73,000 to support various external educational institutions using the UFTR facilities in academic year 2000-01. The actual Reactor Sharing Grant was only in the amount of $40,000, all of which has been well used by the University of Florida as host institution to support various educational institutions in the use of our reactor and associated facilities as indicated in the proposal. These various educational institutions are located primarily within the State of Florida. However, when the 600-mile distance from Pensacola to Miami is considered, it is obvious that this Grant provides access to reactor utilization for a broad geographical region and a diverse set of user institutions serving over fourteen million inhabitants throughout the State of Florida and still others throughout the Southeast.

Advanced Demonstration and Test Reactor Options Study

Energy Technology Data Exchange (ETDEWEB)

Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

Magnetic scanning and interpretation of paleomagnetic data from Prague Synform’s volcanics

Czech Academy of Sciences Publication Activity Database

Kletetschka, Günther; Schnabl, Petr; Šifnerová, Kristýna; Tasáryová, Z.; Manda, Š.; Pruner, Petr

2013-01-01

Roč. 57, č. 1 (2013), s. 103-117 ISSN 0039-3169 R&D Projects: GA ČR GAP210/10/2351 Institutional support: RVO:67985831 Keywords : paleomagnetism * magnetic scanner * magnetic mineralogy * amygdales * magnetic anomalies * magnetic texture * Barrandian Subject RIV: DE - Earth Magnetism, Geodesy, Geography Impact factor: 0.752, year: 2013

The status of the PIK reactor

Energy Technology Data Exchange (ETDEWEB)

Petrov, Yu V [Academy of Sciences of Russia, Petersburg Nuclear Physics Institute, Gatchina, St. Petersburg (Russian Federation)

1992-07-01

This report describes the 100 MW research reactor PIK which is now under construction. The thermal neutron flux in the heavy water reflector exceeds 10{sup 15} cm{sup -2}s{sup -1}; in the light water trap, it is about 4{center_dot}10{sup 15} cm{sup -2}s{sup -1}. The replaceable core vessel allows to vary the parameters of the core over a wide range. The reactor provides sources of hot, cold and ultracold neutrons for 10 horizontal, 6 inclined neutron beams, and 8 neutron guides. At the ends of the beam tubes, the neutron flux is 10{sup 10} - 10{sup 11} cm{sup -2}s{sup -1}. The flux of the long wave neutrons exceeds 10{sup 9} cm{sup -2}s{sup -1}. To ensure precise measurements, the experimental hall is protected against vibrations. The project meets all modern safety requirements. The calculated parameters of the reactor were verified using a full-scale mock-up. Seventy percent of the reactor construction and installation were completed in the beginning of 1992. (author)

Commercial U.S. Vendors Focus on Reducing the Cost of Fast Reactors

International Nuclear Information System (INIS)

Fraser, Dan; Kouhestani, Amir; Parmentola, John; Prince, Robert; Reynolds, Roger

2013-01-01

A US commercial perspective: • Focus on economic benefits of fast reactors (Market Determined): • Compact design, high energy density (modularity); • High temperature output; • Improved energy conversion efficiency; • Process heat application market; • Inherent safety capabilities. • Focus on burn or “breed and burn”; • Helps avoid some non-proliferation challenges. • Focus on reactor alone, not on a re-processing plant; • Reactor is the first place for economic payback; • Happy to burn reprocessed fuel if available

Bylanské sídliště na polykulturní lokalitě v Praze 9 - Černém Mostě

Czech Academy of Sciences Publication Activity Database

Mácalová, Michaela; Frolík, Jan

Misc. 2, - (2011), s. 137-156 ISSN 1211-5169. [Doba popelnicových polí a doba halštatská /11./. Příbram, 07.09.2010-10.09.2010] Institutional research plan: CEZ:AV0Z80020508 Keywords : Bylany culture * Prague * prehistory Subject RIV: AC - Archeology, Anthropology, Ethnology

New results for Palaeozoic volcanic phases in the Prague Basin – magnetic and geochemical studies of Lištice, Czech Republic

Directory of Open Access Journals (Sweden)

Tiiu Elbra

2015-02-01

Full Text Available Palaeo-, rock magnetic and geochemical studies were conducted on volcanic samples from the Lištice area to improve the knowledge of Palaeozoic volcanic evolution in the Prague Basin. The magnetic data display no significant differences between two studied localities, indicating one magnetizing event for both localities. Geochemical data suggest that Lištice basalt could have originated from deep melting of the garnet peridotite mantle source during the attenuation and rifting of the continental lithosphere connected with asthenospheric mantle upwelling. The dataset furthermore supports the evidence of syn- or post-intrusive fluid interactions and low-temperature stages of alteration. The Ti-magnetite within amygdales of the samples was found to be carrying the characteristic remanent magnetization and reflects probably the Permo-Carboniferous remagnetization of volcanic phases.

The advanced MAPLE reactor concept

International Nuclear Information System (INIS)

Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

1989-01-01

High-flux neutron sources are continuing to be of interest both in Canada and internationally to support materials testing for advanced power reactors, new developments in extracted-neutron-beam applications, and commercial production of selected radioisotopes. The advanced MAPLE reactor concept has been developed to meet these needs. The advanced MAPLE reactor is a new tank-type D 2 O reactor that uses rodded low-enrichment uranium fuel in a compact annular core to generate peak thermal-neutron fluxes of 1 x 10 19 n·s -1 in a central irradiation rig with a thermal power output of 50 MW. Capital and incremental development costs are minimized by using MAPLE reactor technology to the greatest extent practicable

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Health physics aspects of advanced reactor licensing reviews

International Nuclear Information System (INIS)

Hinson, C.S.

1995-01-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on open-quotes next-generationclose quotes reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs currently being reviewed by the NRC

Reactor-specific spent fuel discharge projections, 1987-2020

International Nuclear Information System (INIS)

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

50 years of challenging projects and bold visions

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Rataj, J.

2005-01-01

The article commemorates the 50th anniversary of the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, paying particular attention to the department (whose current name is the Department of Nuclear Reactors) that has been educating and training specialists in the theory and technology of nuclear reactors / nuclear power engineering. Both the educational and scientific research activities of the department during the past 15 years are described. The department's fundamental responsibilities include, among other things, the operation and use of the VR-1 training reactor. The prospect of the department in the near future are also outlined. (orig.)

Reactor water level control device

International Nuclear Information System (INIS)

Utagawa, Kazuyuki.

1993-01-01

A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

Training of engineers for Czechoslovak nuclear programme at Czech Technical University in Prague

International Nuclear Information System (INIS)

Klik, F.; Stoll, I.

1983-01-01

Between the year 1959 and the 1970's specialists for the Czechoslovak nuclear programme were only educated at the Faculty of Nuclar and Physics Engineering. In the early 1970's instruction and research related to nuclear power generation was introduced at the mechanical engineering and electrical engineering faculties. The specialization ''Nuclear power facilities'' was introduced within the study field ''Thermal and nuclear machines and equipment'' at the mechanical engineering faculty, and the electrical engineering faculty opened the study course ''Nuclear power plants'' in the study year 1975/1976. Most specialists for the nuclear programme are educated at the Faculty of Nuclear and Physics Engineering in the field ''Nuclear chemical engineering'' and in specializations ''Theory and technology of nuclear reactors'', ''Dosimetry and application of ionizing radiation'' in the study field ''Nuclear engineering''. The Faculty of Nuclear and Physics Engineering also trains specialists in the field ''Structure and materials properties'', the study courses ''Measuring technology'' and ''Control technology'' are run at the electrical engineering faculty and at the mechanical engineering faculty were introduced study courses of ''Applied mechanics'' and ''Mechanical engineering technology''. Graduates of all said study courses may be employed in the nuclear programme. (E.S.)

Reactor use in nuclear engineering programs

International Nuclear Information System (INIS)

Murray, R.L.

1975-01-01

Nuclear reactors for dual use in training and research were established at about 50 universities in the period since 1950, with assistance by the U. S. Atomic Energy Commission and the National Science Foundation. Most of the reactors are in active use for a variety of educational functions--laboratory teaching of undergraduates and graduate students, graduate research, orientation of visitors, and nuclear power plant reactor operator training, along with service to the technical community. As expected, the higher power reactors enjoy a larger average weekly use. Among special programs are reactor sharing and high-school teachers' workshops

Risk Management for Sodium Fast Reactors.

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-01-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

Lessons from early experience in reactor development

International Nuclear Information System (INIS)

Allen, W.

1976-09-01

This paper deals with several issues in U.S. reactor development and demonstration experience. The focus is on the period between 1946 and 1963 during which the Atomic Energy Commission (AEC) guided early reactor research and development (R and D) and conducted the Power Reactor Demonstration Program

Lessons learned from the shut down, planning, and the preparatory activities of decommissioning the research reactor VVR-S Magurele, Bucharest

International Nuclear Information System (INIS)

Dragusin, M.; Copaciu, V.

2006-01-01

The nuclear research reactor type VVR was shut down in December 1997 after forty years of operation. The main characteristics of this reactor are: Thermal power 2 MW, Thermal energy - 9.59 GWhd, Average flux of thermal neutrons-10 13 n/cm 2 .s, nine horizontal channels, sixteen vertical exposure channels, three biological channels, reactor type tank, water used as a moderator, coolant and reflector. The reactor was used in research and radioisotope production. The reactor has been permanently shut down since April 2002, when the decommissioning was officially announced. Discussions regarding funding mechanisms for the conservation phase, and decommissioning (planning, preparatory activities, spent nuclear fuel management), have taken place since five years ago when the final decision of permanent shut down was taken. Quality management includes procedures for recording and archiving the lessons learned. The planning of decommissioning started in 1990 when the reactor was still operational. After fifteen years the regulatory body has not yet approved the decommissioning plan for the reactor. In this paper the following aspects are discussed: decommissioning strategy from safe enclosure to immediate dismantling, specific features of the site (treatment of radioactive waste near reactor) and state of decommissioning, use of the lessons learned in the planning of decommissioning for the other two small nuclear facilities situated in the same area with VVR-reactor: Sub critical Assembly 'HELEN' and Zero Power Critical Reactor RP-0, AFR ponds for spent nuclear fuel, other radiological facilities for radioisotopes production facilities radiation processing and accelerators. Preparatory activities for decommissioning have included: elaboration of a plan (inter alia, justification of the selected strategy, management of the radioactive waste in accordance with the waste acceptance criteria), reactor storage in parallel with the removal of the equipment and materials used in

Fuel assembly for FBR type reactor and reactor core thereof

International Nuclear Information System (INIS)

Kobayashi, Kaoru.

1998-01-01

The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

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

Reactor container cooling device

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1995-11-10

The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

The Jules Horowitz reactor project, a driver for revival of the research reactor community

Energy Technology Data Exchange (ETDEWEB)

Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

2010-07-01

The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

2016-04-17

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.