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Sample records for reaktore azmayeshgahi-ye jabejai

  1. Fire modeling of the Heiss Dampf Reaktor containment

    SciTech Connect

    Nicolette, V.F.; Yang, K.T.

    1995-09-01

    This report summarizes Sandia National Laboratories` participation in the fire modeling activities for the German Heiss Dampf Reaktor (HDR) containment building, under the sponsorship of the United States Nuclear Regulatory Commission. The purpose of this report is twofold: (1) to summarize Sandia`s participation in the HDR fire modeling efforts and (2) to summarize the results of the international fire modeling community involved in modeling the HDR fire tests. Additional comments, on the state of fire modeling and trends in the international fire modeling community are also included. It is noted that, although the trend internationally in fire modeling is toward the development of the more complex fire field models, each type of fire model has something to contribute to the understanding of fires in nuclear power plants.

  2. Combined nitrification/denitrification in a membrane reactor.

    PubMed

    Walter, B; Haase, C; Räbiger, N

    2005-08-01

    An ever stricter legislation regulating wastewater leads to an increasing demand for biological treatment plants which are able to selectively eliminate nitrogen from wastewaters with a high influent concentration, even when operating in partial influent mode. A membrane-tube-module (MSM) reactor (Membran-Schlauch-Modul-Reaktor) was constructed and realized in the IUV at the University of Bremen. The present approach makes use of all the various layers of the whole biofilm, enabling nitrification and denitrification processes to run simultaneously in one and the same biofilm under optimized conditions. The biological degradation capacity of the system was first successfully tested with synthetic wastewater, and subsequently in a real application with effluents from a recycling of animal carcasses plant and from a coke-oven plant. A mathematical model was devised which describes this biofilm system. The resulting equations were solved by means of the simulation software AQUASIM.

  3. Fuel Element Transfer Cask Modelling Using MCNP Technique

    SciTech Connect

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-05

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  4. Fuel Element Transfer Cask Modelling Using MCNP Technique

    NASA Astrophysics Data System (ADS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  5. Current Status of VHTR Technology Development

    SciTech Connect

    David Petti; Hans Gougar; Richard Wright; William Windes; Steve Herring; Richard Schultz; Paul Humrickhouse

    2010-10-01

    Abstract – High Temperature Gas-cooled Reactors (HTGRs) featuring particle fuel reached the stage of commercial deployment in the mid-1980s with the Fort St.Vrain and Thorium HochTemperatur Reaktor feeding electricity to the grids in the United States and West Germany, respectively. The technology was then adopted by Japan and China with the operation of the High Temperature Test Reactor in Oarai, Japan and the High Temperature Reactor (HTR-10) in China. Increasing the outlet temperature of the HTGR to even higher temperatures above 900°C will improve the thermodynamic efficiency of the system and enable application of a new class of gas reactor, the very high temperature reactor, to provide process heat, electricity, and hydrogen to chemical industries with the attendant benefits of improved energy security and reduced CO2 emissions. However, the increase in coolant outlet temperature presents a number of technical challenges associated with fuel, materials, power conversion, and analysis methods for the reactor and hydrogen production. The U.S. Department of Energy is sponsoring a broad program of research and development with a goal of addressing the technical challenges over a broad range of outlet temperatures as part of the Next Generation Nuclear Plant Project. This paper describes the research and development activities that are currently underway to realize the technologies needed for an HTGR that features outlet temperatures of 750 to 950°C.

  6. Radioactivity of spent TRIGA fuel

    SciTech Connect

    Usang, M. D. Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  7. Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

    SciTech Connect

    Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D

    2003-08-15

    The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy.

  8. Numerical Simulation of a Compartment Fire in a Nuclear Power Plant Containment Building

    SciTech Connect

    Jason Floyd

    2002-07-01

    The current trend towards the increased use of risk assessment in the regulation of nuclear power plants will inevitably result in changes in the analysis of fire protection systems and the methods of analysis. Before fire protection can be regulated on a risk basis, a consensus must be reached on a number of issues. One key issue is what types of computational tools will be allowable for analyzing fire events, and what types of scenarios those tools will be approved for use. Reaching this consensus will require an understanding of the types of computational tools available and their inherent advantages and disadvantages. To aid with this understanding, three different methods of fire simulation are applied to an oil pool fire test in the HDR (Heiss Dampf Reaktor) containment test facility. These methods are a hand calculation, the zone model code CFAST (Consolidated Model of Fire Growth and Smoke Transport), and the computational fluid dynamics code FDS (Fire Dynamics Simulator). Each is applied to a steady-state portion of the test using, to the extent possible, the same set of input parameters. The results of the computation are compared to the test data. The comparisons show that each method is potentially suitable for use depending on the information required from the simulation. Each method will potentially have a role to play in risk based regulation depending on the scenario. (authors)

  9. The evolution of the break preclusion concept for nuclear power plants in Germany

    SciTech Connect

    Schulz, H.

    1997-04-01

    In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A = open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.

  10. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  11. Radioactivity of spent TRIGA fuel

    NASA Astrophysics Data System (ADS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  12. (HFR-B1 experiment reporting and capsule disassembly)

    SciTech Connect

    Myers, B.F.

    1991-02-22

    The traveler visited the Joint Research Centre (JRC), Petten, The Netherlands, the Forschungszentrum GmbH (KFA), Juelich, Germany; and the Zentralinstitut fuer Kernforschung (ZfK), Rossendorf, Germany, during the period January 28 through February 9. At JRC, the analysis of the experiment HFR-B1 was discussed; a new schedule for issuance of the final data report was established. Other discussions at JRC concerned the capabilities of Petten to conduct two reactor experiments being proposed under the US/FRG cooperative program and the initial results of a proof test of Germany fuel spheres. At KFA, the main emphasis was on the disassembly of capsules 2 and 3 of the HFR-B1 experiment and agreement on the examinations and tests to be conducted with the disassembled components. The disassembly of capsule 3 was observed. Extensive discussions were conducted on the work, both experimental and analytical, being conducted in the Institut fuer Sicherheitsforschung und Reaktor Technologie. A major portion of the experimental work is being conducted at ZfK and a visit to this laboratory, sponosored by the KFA, was made on February 6 and 7. Cooperation with the US on the experimental and analytical work in the safety area was strongly emphasized. 1 tab.

  13. Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Masri Husam Fayiz, Al

    2017-01-01

    The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

  14. MHTGR [modular high-temperature gas-cooled reactor] core physics validation plan

    SciTech Connect

    Baxter, A.; Hackney, R.

    1988-01-01

    This document contains the verification and validation (V&V) plan for analytical methods utilized in the nuclear design for normal and off-normal conditions within the Modular High-Temperature Gas-Cooled Reactor (MHTGR). Regulations, regulatory guides, and industry standards have been reviewed and the approach for V&V has been developed. MHTGR core physics methods are described and the status of previous V&V is summarized within this document. Additional work required to verify and validate these methods is identified. The additional validation work includes comparison of calculations with available experimental data, benchmark comparison of calculations with available experimental data, benchmark comparisons with other validated codes, results from a cooperative program now underway at the Arbeitsgemeinschaft Versuchs-Reaktor GmbH (AVR) facility in Germany, results from a planned series of experiments on the Compact Nuclear Power Source (CNPS) facility at Los Alamos, and detailed documentation of all V&V studies. In addition, information will be obtained from planned international cooperative agreements to provide supplemental data for V&V. The regulatory technology development plan will be revised to include these additional experiments. A work schedule and cost estimate for completing this plan is also provided. This work schedule indicates the timeframe in which major milestones must be performed in order to complete V&V tasks prior to the issuance of preliminary design approval from the NRC. The cost to complete V&V tasks for core physics computational methods is estimated to be $2.2M. 41 refs., 13 figs., 8 tabs.

  15. Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

    SciTech Connect

    Weigl, M.

    2012-07-01

    Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactor with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not possible

  16. Photocatalitic Properties of Tio2 and ZnO Nanopowders / Tio2 un Zno Nanopulveru Fotokatalitiskās Īpašības

    NASA Astrophysics Data System (ADS)

    Grigorjeva, L.; Rikveilis, J.; Grabis, J.; Jankovica, Dz.; Monty, C.; Millers, D.; Smits, K.

    2013-08-01

    Photocatalytic activity of TiO2 and ZnO nanopowders is studied depending on the morphology, grain sizes and method of synthesizing. Photocatalysis of the prepared powders was evaluated by degradation of the methylene blue aqueous solution. Absorbance spectra (190-100 nm) were measured during exposure of the solution to UV light. The relationships between the photocatalytic activity and the particle size, crystal polymorph phases and grain morphology were analyzed. The photocatalytic activity of prepared TiO2 nanopowders has been found to depend of the anatase-to-rutile phase ratio. Comparison is given for the photocatalytic activity of ZnO nanopowders prepared by sol-gel and solar physical vapour deposition (SPVD) methods Darbā pētīta fotokatalīzes efektivitāte ar dažādām metodēm sintezētiem TiO2 and ZnO nanopulveriem, kuriem ir atšķirīga morfoloģija un grauda izmērs. Foto katalīzes process raksturots ar metilenzilā sagraušanu ūdens šķīdumā, to apstarojot ar UV gaismu. Analizēta fotokatalīzes efektivitātes atkarība no grauda izmēra, nanokristālu graudu morfoloģijas, TiO2 nanopulveru anatasa-rutīla fāžu svara attiecībām. Parādīts, ka fotokatalītiskā efektivitāte ir atšķirīga TiO2 nanopulveriem sintezētiem ar dažādām metodēm: sola-gēla un tvaicēšanu-kondensēšanu saules reaktorā. Salīdzināta fotokatalīzes efektivitāte ZnO un TiO2 nanopulveriem un secināts, ka ZnO nanopulveri ar tetrapodu morfoloģiju ir labs fotokatalizators

  17. Influence of Light Intensity and Temperature on Cultivation of Microalgae Desmodesmus Communis in Flasks and Laboratory-Scale Stirred Tank Photobioreactor

    NASA Astrophysics Data System (ADS)

    Vanags, J.; Kunga, L.; Dubencovs, K.; Galvanauskas, V.; Grīgs, O.

    2015-04-01

    Optimization of the microalgae cultivation process and of the bioprocess in general traditionally starts with cultivation experiments in flasks. Then the scale-up follows, when the process from flasks is transferred into a laboratory-scale bioreactor, in which further experiments are performed before developing the process in a pilot-scale reactor. This research was done in order to scale-up the process from a 0.4 1 shake flask to a 4.0 1 laboratory-scale stirred-tank photobioreactor for the cultivation of Desmodesmus (D.) communis microalgae. First, the effect of variation in temperature (21-29 ºC) and in light intensity (200-600 μmol m-2s-1) was studied in the shake-flask experiments. It was shown that the best results (the maximum biomass concentration of 2.72 g 1-1 with a specific growth rate of 0.65 g g-1d-1) can be achieved at the cultivation temperature and light intensity being 25 °C and 300 μmol m2s-1, respectively. At the same time, D. communis cultivation under the same conditions in stirred-tank photobioreactor resulted in average volumetric productivities of biomass due to the light limitation even when the light intensity was increased during the experiment (the maximum biomass productivity 0.25 g 1-1d-1; the maximum biomass concentration 1.78 g 1-1). Mikroaļģu kultivēšanas procesa optimizēšana parasti sākas ar kultivēšanas eksperimentiem kolbās. Tālāk seko procesa pārnese uz laboratorijas mēroga fotobioreaktoru, kurā tiek veikti tālāki eksperimenti, pirms tiek izveidots pilota mēroga reaktors. Šis pētījums tika veikts ar mērķi, pārnest Desmodesmus communis kultivēšanas procesu no 0.4 1 kolbas uz 4.0 1 laboratorijas fotobioreaktoru. Vispirms tika pētīta dažādu temperatūru (21-29 ºC) un gaismas intensitātes (200-600 μmol m-2s-1) ietekme uz aļģu biomasu veicot eksperimentus kolbās. Labākie rezultāti (maksimālā biomasas koncentrācija 2.72 g 1-1; īpatnējais augšanas ātrums 0.65 g g-1d-1) sasniegti, kad

  18. Properties of Waste from Coal Gasification in Entrained Flow Reactors in the Aspect of Their Use in Mining Technology / Właściwości odpadów ze zgazowania węgla w reaktorach dyspersyjnych w aspekcie ich wykorzystania w technologiach górniczych

    NASA Astrophysics Data System (ADS)

    Pomykała, Radosław

    2013-06-01

    Most of the coal gasification plants based of one of the three main types of reactors: fixed bed, fluidized bed or entrained flow. In recent years, the last ones, which works as "slagging" reactors (due to the form of generated waste), are very popular among commercial installations. The article discusses the characteristics of the waste from coal gasification in entrained flow reactors, obtained from three foreign installations. The studies was conducted in terms of the possibilities of use these wastes in mining technologies, characteristic for Polish underground coal mines. The results were compared with the requirements of Polish Standards for the materials used in hydraulic backfill as well as suspension technology: solidification backfill and mixtures for gob caulking. Większość przemysłowych instalacji zgazowania węgla pracuje w oparciu o jeden z trzech głównych typów reaktorów: ze złożem stałym, dyspersyjny lub fluidalny. W zależności od rodzaju reaktora oraz szczegółowych rozwiązań instalacji, powstające uboczne produkty zgazowania mogą mieć różną postać. Zależy ona w dużej mierze od stosunku temperatury pracy reaktora do temperatury topnienia części mineralnych zawartych w paliwie, czyli do temperatury mięknienia i topnienia popiołu. W ostatnich latach bardzo dużą popularność wśród instalacji komercyjnych zdobywają reaktory dyspersyjne "żużlujące". W takich instalacjach żużel jest wychwytywany i studzony po wypłynięciu z reaktora. W niektórych przypadkach oprócz żużla powstaje jeszcze popiół lotny, wychwytywany w systemach odprowadzania spalin. Może być on pozyskiwany oddzielnie lub też zawracany do komory reaktora, gdzie ulega stopieniu. Wszystkie z analizowanych odpadów - trzy żużle oraz popiół pochodzą właśnie z tego typu instalacji. Tylko z jednej z nich pozyskano zarówno żużel jak i popiół, z pozostałych dwóch jedynie żużel. Odpady te powstały, jako uboczny produkt zgazowania w