WorldWideScience

Sample records for experimental reactor iter

  1. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-18

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  2. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  3. Analysis of quench-vent pressures for present design of ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coils

    Energy Technology Data Exchange (ETDEWEB)

    Slack, D.S.

    1989-09-20

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three- dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 1 fig., 3 tabs.

  4. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers & Constructors and Chicago Bridge & Iron (Raytheon/CB&I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB&I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB&I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB&I and documented accordingly.

  5. Recent progress on the hydrogen storage properties of ZrCo-based alloys applied in International Thermonuclear Experimental Reactor (ITER

    Directory of Open Access Journals (Sweden)

    Feng Wang

    2017-02-01

    Full Text Available Under the development of International Thermonuclear Experimental Reactor (ITER system aimed at realizing the controllable fusion reaction to solve the energy crisis fundamentally, there is an urgent need to find an appropriate material for tritium handling. ZrCo alloy is considered to be a promising candidate for the storage and delivery of hydrogen isotopes due to the favorable characteristics such as low plateau pressure for absorption, high dissociation pressure at moderate temperature and better ability of trapping 3He. However, the hydrogen induced disproportionation and the slower recovery/deliverty rate of ZrCo-based alloys have limited their further application in ITER system. This review summarizes the efforts towards enhancing the hydrogen storage properties of ZrCo-based alloys including element substitution, surface modification, disproportionation mechanism investigation and the isotope effect study. Element substitution and surface modification play positive role to improve the anti-disproportionation ability and kinetic property of the alloys. However, the ZrCo-based alloys require to be further modified by more attempts such as new composition, novelty modification method or catalyst addition in order to better satisfy the application demands for tritium handling. Moreover, new insight for further understanding the inner disproportionation mechanisms of this material is needed by combining the advance characterization and theoretical analysis, which is in favor of addressing the disproportionation problem of the ZrCo-based alloys essentially.

  6. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka , Ibaraki (Japan)

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  7. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Document Server

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  8. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  9. Progress on ITER remote experimentation centre

    Energy Technology Data Exchange (ETDEWEB)

    Ozeki, Takahisa, E-mail: ozeki.takahisa@jaea.go.jp [Japan Atomic Energy Agency, 2-166 Obuchi Rokkasho, Kitakami-gun, Aomori 039-3212 (Japan); Clement-Lorenzo, Susana [Fusion for Energy, Torres Diagonal Litoral, B3, 13/03, Barcelona 08019 (Spain); Nakajima, Noriyoshi [National institute for Fusion Science and Project leader of IFERC, 2-166 Obuchi, Rokkasho, Kamikita-gun, Aomori 039-3212 (Japan)

    2016-11-15

    Construction of ITER remote experimentation centre (REC) based on the broader approach (BA) activity of the joint program of Japan and Europe (EU) is progressing. In order to make the future experiments of ITER and JT-60SA effectively and efficiently implemented, development of a remote experiment system by using the Satellite Tokamak (JT-60SA) facilities was planned and the development of software for the remote experiment is ongoing, including the systems for the remote connection and the communication between the remote site and the on-site facility. The network system from REC in Rokkasho-site of Japan to the network in EU was established in collaboration with the National Institute of Informatics (NII). Effective data transfer method that is capable of fast transfer speeds in the gigabit range is investigated. Data transfer at the rate of several Gbps was successfully obtained between the institutes in Japan. The preliminary versions of the software for data analysis are developed, such as for visualization of time dependent experimental data and transport simulations, visualization of plasma boundary/equilibrium and spatial profiles of diagnostic data. The remote data access program and an integrated platform for Documentation and Experiment Management are also being developed. A remote experiment room in the Rokkasho-site in Japan was designed and the construction started. The function of REC will be tested and the total system will be demonstrated by the middle of 2017.

  10. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id [Department of Nuclear Physics, Faculty of Mathematic and Natural Sciences, Institut Teknologi Bandung (Indonesia); Yazid, Putranto Ilham [Research and Development of Nuclear Association (Indonesia)

    2015-09-30

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  11. Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

    Science.gov (United States)

    Indah Rosidah, M.; Suud, Zaki; Yazid, Putranto Ilham

    2015-09-01

    The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With the tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature

  12. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Odegard, B.C. Jr.; Cadden, C.H. [Sandia National Labs., Livermore, CA (United States); Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Slattery, K.T. [Boeing Co., St. Louis, MO (United States)

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report.

  13. The neutronic analysis of opportunity of ITER blanket element tests in RF research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lopatkin, A.; Tocheniy, L. [ENTEK-RDIPE, Moscow (Russian Federation)

    1994-12-31

    In the framework of development of plan of in-pile radiative tests of ITER blanket elements the calculations are carried out of the models of tritium-producing elements in loop channels, placed in the number of Russian various type test reactors. There are presented: (1) The variants of models of blanket, on the base of which the set of experiment goal parameters and its ranges are formed; (2) Outline of loop channel; (3) The experimental opportunities of research reactors with thermal (SM-3, MIR, IVV-2M, RBT) and fast (BOR-60, EBR) spectra of neutrons; (4) The calculation procedures - settlement models, codes. The results are given: (1) power generation rates in components of channel; (2) the tritium breeding rate; (3) the helium production rate in beryllium; (4) the neutron group fluxes; (5) absorption rates in zones of loop channel. The possible reactivity effects due to experimental channel accommodation in reactor core and to radiated sample replace inside of the channel are shown. The last section includes the recommendations for the choice of reactor acceptable from the neutronics point of view, and for the next study directions and stages.

  14. Study of heat and synchrotron radiation transport in fusion tokamak plasmas. Application to the modelling of steady state and fast burn termination scenarios for the international experimental fusion reactor ITER

    Energy Technology Data Exchange (ETDEWEB)

    Villar Colome, J. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Universitat Polytechnica de Catalunya (Spain)

    1997-12-01

    The aim of this thesis is to give a global scope of the problem of energy transport within a thermonuclear plasma in the context of its power balance and the implications when modelling ITER operating scenarios. This is made in two phases. First, by furnishing new elements to the existing models of heat and synchrotron radiation transport in a thermonuclear plasma. Second, by applying the improved models to plasma engineering studies of ITER operating scenarios. The scenarios modelled are the steady state operating point and the transient that appears to have the biggest technological implications: the fast burn termination. The conduction-convection losses are modelled through the energy confinement time. This parameter is empirically obtained from the existing experimental data, since the underlying mechanisms are not well understood. In chapter 2 an expression for the energy confinement time is semi-analytically deduced from the Rebut-Lallia-Watkins local transport model. The current estimates of the synchrotron radiation losses are made with expressions of the dimensionless transparency factor deduced from a 0-dimensional cylindrical model proposed by Trubnikov in 1979. In chapter 3 realistic hypothesis for the cases of cylindrical and toroidal geometry are included in the model to deduce compact explicit expressions for the fast numerical computation of the synchrotron radiation losses. Numerical applications are provided for the cylindrical case. The results are checked against the existing models. In chapter 4, the nominal operating point of ITER and its thermal stability is studied by means of a 0-dimensional burn model of the thermonuclear plasma in ignition. This model is deduced by the elements furnished by the plasma particle and power balance. Possible heat overloading on the plasma facing components may provoke severe structural damage, implying potential safety problems related to tritium inventory and metal activation. In chapter 5, the assessment

  15. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  16. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  17. Experimental Vertical Stability Studies for ITER Performance and Design Guidance

    Energy Technology Data Exchange (ETDEWEB)

    Humphreys, D A; Casper, T A; Eidietis, N; Ferrera, M; Gates, D A; Hutchinson, I H; Jackson, G L; Kolemen, E; Leuer, J A; Lister, J; LoDestro, L L; Meyer, W H; Pearlstein, L D; Sartori, F; Walker, M L; Welander, A S; Wolfe, S M

    2008-10-13

    Operating experimental devices have provided key inputs to the design process for ITER axisymmetric control. In particular, experiments have quantified controllability and robustness requirements in the presence of realistic noise and disturbance environments, which are difficult or impossible to characterize with modeling and simulation alone. This kind of information is particularly critical for ITER vertical control, which poses some of the highest demands on poloidal field system performance, since the consequences of loss of vertical control can be very severe. The present work describes results of multi-machine studies performed under a joint ITPA experiment on fundamental vertical control performance and controllability limits. We present experimental results from Alcator C-Mod, DIII-D, NSTX, TCV, and JET, along with analysis of these data to provide vertical control performance guidance to ITER. Useful metrics to quantify this control performance include the stability margin and maximum controllable vertical displacement. Theoretical analysis of the maximum controllable vertical displacement suggests effective approaches to improving performance in terms of this metric, with implications for ITER design modifications. Typical levels of noise in the vertical position measurement which can challenge the vertical control loop are assessed and analyzed.

  18. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  19. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  20. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Osinga, T. [ETH-Zuerich (Switzerland); Olalde, G. [CNRS Odeillo (France); Steinfeld, A. [PSI and ETHZ (Switzerland)

    2005-03-01

    A numerical model is formulated for the SOLZINC solar chemical reactor for the production of Zn by carbothermal reduction of ZnO. The model involves solving, by the finite-volume technique, a 1D unsteady state energy equation that couples heat transfer to the chemical kinetics for a shrinking packed bed exposed to thermal radiation. Validation is accomplished by comparison with experimentally measured temperature profiles and Zn production rates as a function of time, obtained for a 5-kW solar reactor tested at PSI's solar furnace. (author)

  1. Validation of the inspections with ultrasound of the welds of the reactor of ITER vacuum vessel; Validacion de las inspecciones con ultrasonidos de las soldaduras de la Vasija de Vacio del reactor del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, A.; Fernandez, F.; Perez, C.; Sillero, J. A.

    2013-07-01

    The ITER fusion reactor vacuum vessel has thousands of welding austenitic with shapes and different manufacturing processes. The RCC-MR code, which is that applied to the manufacture of the fusion reactor, requires a volumetric test all of them. This test should be mainly by x-rays and welds where it was not possible to use this method, ultrasonic.09-06.

  2. Burn control of an ITER-like fusion reactor using fuzzy logic

    Science.gov (United States)

    Garcia-Amador, A. Sair; Martinell, Julio J.

    2016-10-01

    The fuel burn in a fusion reactor has to be kept at a nearly constant rate in order to have a steady power exhaust. Here, we develop a control system based on a fuzzy logic controller in order that adjusts external parameters to keep the plasma temperature and density at the design values of a reactor of the characteristics of ITER. The control parameters chosen are the D-T refueling rate, the auxiliary heating power and a neutral helium beam. We use a fuzzy controller of the Mamdani type that uses a number of membership functions appropriate to produce a response to parameter deviations that minimizes the response time. The inference rules are determined in a way to provide stabilization to all perturbations of the temperature, density and alpha particle fraction. The dynamical response of the reactor is simulated with a 0D model that uses confinement times provided by the ITER scaling. We show that the system is feedback stabilized for a large range of parameters around the nominal values. The recovery time after a departure from the steady values is of the order of one second. We compare the results with another control system based on neural networks that was developed previously. Funded by projects PAPIIT IN109115 and Conacyt 152905.

  3. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  4. Economic Impacts on the United States of Siting Decisions for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Peerenboom, J. P.; Hanson, M. E.; Huddleston, J. R.; Wolsko, T. D.

    1997-12-01

    This paper presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  5. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  6. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri

    2008-01-01

    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  7. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  8. PARALLEL ITERATIVE RECONSTRUCTION OF PHANTOM CATPHAN ON EXPERIMENTAL DATA

    Directory of Open Access Journals (Sweden)

    M. A. Mirzavand

    2016-01-01

    Full Text Available The principles of fast parallel iterative algorithms based on the use of graphics accelerators and OpenGL library are considered in the paper. The proposed approach provides simultaneous minimization of the residuals of the desired solution and total variation of the reconstructed three- dimensional image. The number of necessary input data, i. e. conical X-ray projections, can be reduced several times. It means in a corresponding number of times the possibility to reduce radiation exposure to the patient. At the same time maintain the necessary contrast and spatial resolution of threedimensional image of the patient. Heuristic iterative algorithm can be used as an alternative to the well-known three-dimensional Feldkamp algorithm.

  9. Installation of the ITER committee industry. Participants guide; Installation du Comite industrie ITER. Dossier des participants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    ITER is an international project to design and build an experimental fusion reactor based on the tokamak concept. This guide presents the ITER project and objectives and the associated organizations in France, the recommendations and actions for ITER, the industrial mobilization, the industrial committee and its members, technological sheets for the enterprises and the statistical document of the SESSI. (A.L.B.)

  10. Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO

    Science.gov (United States)

    Rubel, M.; Moon, Soonwoo; Petersson, P.; Garcia-Carrasco, A.; Hallén, A.; Krawczynska, A.; Fortuna-Zaleśna, E.; Gilbert, M.; Płociński, T.; Widdowson, A.; Contributors, JET

    2017-12-01

    Optical spectroscopy and imaging diagnostics in next-step fusion devices will rely on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and in laboratory systems. This work deals with comprehensive tests of mirrors: (a) exposed in JET with the ITER-like wall (JET-ILW); (b) irradiated by hydrogen, helium and heavy ions to simulate transmutation effects and damage which may be induced by neutrons under reactor conditions. The emphasis has been on surface modification: deposited layers on JET mirrors from the divertor and on near-surface damage in ion-irradiated targets. Analyses performed with ion beams, microscopy and spectro-photometry techniques have revealed: (i) the formation of multiple co-deposited layers; (ii) flaking-off of the layers already in the tokamak, despite the small thickness (130–200 nm) of the granular deposits; (iii) deposition of dust particles (0.2–5 μm, 300–400 mm‑2) composed mainly of tungsten and nickel; (iv) that the stepwise irradiation of up to 30 dpa by heavy ions (Mo, Zr or Nb) caused only small changes in the optical performance, in some cases even improving reflectivity due to the removal of the surface oxide layer; (v) significant reflectivity degradation related to bubble formation caused by the irradiation with He and H ions.

  11. Experimental modelling of plasma-graphite surface interaction in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Martynenko, Yu.V.; Guseva, M.I.; Gureev, V.M.; Danelyan, L.S.; Neumoin, V.E.; Petrov, V.B.; Khripunov, B.I.; Sokolov, Yu.A.; Stativkina, O.V.; Stolyarova, V.G. [Rossijskij Nauchnyj Tsentr ``Kurchatovskij Inst.``, Moscow (Russian Federation); Vasiliev, V.I.; Strunnikov, V.M. [TRINITI, Troizk (Russian Federation)

    1998-10-01

    The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10{sup 22} cm{sup -2} and subsequent irradiation by power (250 MW/cm{sup 2}) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770 C and 1150 C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength. (orig.) 9 refs.

  12. Experimental validation of calculations of decay heat induced by 14 MeV neutron activation of ITER materials

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, N.P.; Bartels, H.-W.; Saji, G. [ITER Joint Central Team, San Diego, La Jolla (United States); Ikeda, Y.; Maekawa, F. [Japan Atomic Energy Research Institute, Tokai-mura (Japan); Cambi, G. [Physics Department, Bologna University, Bologna (Italy); Cepraga, D.G. [ENEA, INN-FIS-MACO, Bologna (Italy); Cheng, E.T. [TSI Research Inc., Solana Beach, CA (United States); Forrest, R.A.; Sublet, J.-C. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Iida, H.; Santoro, R.T. [ITER Joint Central Team, Garching (Germany); Khater, H.Y.; Sawan, M.E. [Fusion Technology Institute, University of Wisconsin, Madison, WI (United States); Youssef, M.Z. [Mechanical and Aerospace Engineering Department, University of California at Los Angeles, Los Angeles, CA (United States)

    1999-03-01

    In a D-T fusion tokamak, neutron activation of structural material will result in a source of heat from radioactive decay after shutdown of the plasma. Although relatively small, in certain postulated loss-of-cooling accidents this decay heat could drive a temperature transient of modest proportions. In safety studies of the International Thermonuclear Experimental Reactor (ITER), such accident scenarios are analysed for their possible consequences, requiring an accurate knowledge of the level and time-dependence of the decay heat in components in which it is significant. This is calculated by activation modelling using computer codes with libraries of nuclear data. This paper describes an international benchmark activity to validate several of these codes and data, using the results of direct measurements of decay heat in ITER-relevant materials after exposure to a flux of 14 MeV neutrons. The results show consistency between the codes and, when using data from the recently-compiled FENDL/A-2 activation data library, good agreement with the experiments over the time-scales of interest in ITER safety studies. (orig.) 24 refs.

  13. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.

    1976-01-01

    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  14. Upgrading program of the experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, A.; Yogo, S. [Japan Nuclear Cycle Development Institute, Iibaraki-Ken (Japan)

    2001-07-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  15. Investigation of mixing chamber for experimental FGD reactor

    Directory of Open Access Journals (Sweden)

    Novosád Jan

    2017-01-01

    Full Text Available This article deals with numerical investigation of flow and mixing of air and sulphur dioxide SO2 in designated mixing chamber. The mixing chamber is a part of experimental laboratory reactor designed for simulating the flue gas desulfurization (FGD process. Aim of this work is the numerical investigation of effect of different mixing chamber geometries to mixture composition, especially to mass fraction of sulphur dioxide. Using of similar concentration of sulphur dioxide in the experimental reactor as in the real process is necessary to be able to make additional research. Conclusion describes the effect of different geometries of mixing chamber to mixing. The aim of this work is to develop perfectly works mixing chamber, which will be manufactured and then implemented into experimental FGD reactor. The results will be validated by experiment after the mixing chamber will be manufactured.

  16. Status of experimental data related to Be in ITER materials R and D data bank

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Shigeru [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    To keep traceability of many valuable raw data that were experimentally obtained in the ITER Technology R and D Tasks related to materials for In-Vessel components (divertor, first wall, blanket, vacuum vessel, etc.) and to easily make the best use of these data in the ITER design activities, the `ITER Materials R and D Data Bank` has been built up, with the use of Excel{sup TM} spread sheets. The paper describes status of experimental data collected in this data bank on thermo-mechanical properties of unirradiated and neutron irradiated Be, on plasma-material interactions of Be, on mechanical properties of various kinds of Be/Cu joints (including plasma sprayed Be), and on thermal fatigue tests of Be/Cu mock-ups. (author)

  17. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  18. Experimental and design activities on WDS and ISS as EU contribution to ITER fue

    Energy Technology Data Exchange (ETDEWEB)

    Cristescu, Ion; Cristescu, Iona; Glugla, M.; Hellriegel, G.; Michling, R.; Schaefer, P.; Welte, S.; Wurster, W.; Doerr, L. [Forschungszentrum Karlsruhe (Germany); Murdoch, D. [EFDA Garching (Germany)

    2007-07-01

    The Water Detritiation System (WDS) of ITER is one of the key systems to control the tritium content in the effluents streams, to recover as much tritium as possible and consequently to minimize the impact on the environment. In order to mitigate the concern over tritium releases into the environment during pulsed operation of the Torus, the WDS and Isotope Separation System (ISS) will be operated in such way that WDS will be a final barrier for the processed protium waste gas stream discharged from ISS. The ITER ISS consists of a cascade of four cryogenic distillation columns with the aim to process mainly two gas streams, one from Torus Exhaust Processing (TEP) and other from WDS mixed with the returned stream from Neutral Beam Injectors (NBI). The behaviour of the CD cascade has to be characterized with high accuracy in view of thermal and isotopic fluctuations during Torus pulses. To support the research activities needed to characterize the performances of various components for WDS and ISS processes in various working conditions and configurations as needed for ITER detailed design, an experimental facility called TRENTA based on the combination Combined Electrolysis Catalytic Exchange (CECE) - Cryogenic Distillation (CD), representative of the ITER WDS and ISS protium separation column is in operation at TLK. The CECE process consists of a solid polymer electrolyser unit as envisaged to be used in ITER WDS, and an 8 m Liquid Phase Catalytic Exchange Column (LPCE). The CD system consists of a refrigeration unit of 250 W cooling capacity at 16 K and a cryogenic distillation column of 55 mm in diameter and 2.7 m as process lengths. The experimental program on the TRENTA facility is focused on two major issues: - To investigate the separation performances and liquid hold up of different packings potentially to be used on cryogenic distillation process and the separation performances of the catalyst/packing for the LPCE process. - Trade off studies between the

  19. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40 %) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2–3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  20. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  1. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  2. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  3. Current status of the European contribution to the Remote Data Access System of the ITER Remote Experimentation Centre

    Energy Technology Data Exchange (ETDEWEB)

    De Tommasi, G., E-mail: detommas@unina.it [Fusion for Energy, 08019 Barcelona (Spain); Consorzio CREATE/DIETI, Università degli Studi di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Manduchi, G. [Consorzio RFX, Corso Stati Uniti 4, Padova 35127 (Italy); Muir, D.G. [CCFE, Culham Science Centre, Abingdon OX14 3EA (United Kingdom); Ide, S.; Naito, O.; Urano, H. [Japan Atomic Energy Agency, Naka Fusion Institute, Naka, Ibaraki 311-0193 (Japan); Clement-Lorenzo, S. [Fusion for Energy, 08019 Barcelona (Spain); Nakajima, N. [BA IFERC Project Team, Rokkasho-mura, Aomori 039-3212 (Japan); Ozeki, T. [Japan Atomic Energy Agency, Naka Fusion Institute, Naka, Ibaraki 311-0193 (Japan); Sartori, F. [Fusion for Energy, 08019 Barcelona (Spain)

    2015-10-15

    The ITER Remote Experimentation Centre (REC) is one of the projects under implementation within the BA agreement. The final objective of the REC is to allow researchers to take part in the experimentation on ITER from a remote location. Before ITER first operations, the REC will be used to evaluate ITER-relevant technologies for remote participation. Among the different software tools needed for remote participation, an important one is the Remote Data Access System (RDA), which provides a single software infrastructure to access data stored at the remotely participating experiment, regardless of the geographical location of the users. This paper introduces the European contribution to the RDA system for the REC.

  4. Two conceptual designs of helical fusion reactor FFHR-d1A based on ITER technologies and challenging ideas

    Science.gov (United States)

    Sagara, A.; Miyazawa, J.; Tamura, H.; Tanaka, T.; Goto, T.; Yanagi, N.; Sakamoto, R.; Masuzaki, S.; Ohtani, H.; The FFHR Design Group

    2017-08-01

    The Fusion Engineering Research Project (FERP) at the National Institute for Fusion Science (NIFS) is conducting conceptual design activities for the LHD-type helical fusion reactor FFHR-d1A. This paper newly defines two design options, ‘basic’ and ‘challenging.’ Conservative technologies, including those that will be demonstrated in ITER, are chosen in the basic option in which two helical coils are made of continuously wound cable-in-conduit superconductors of Nb3Sn strands, the divertor is composed of water-cooled tungsten monoblocks, and the blanket is composed of water-cooled ceramic breeders. In contrast, new ideas that would possibly be beneficial for making the reactor design more attractive are boldly included in the challenging option in which the helical coils are wound by connecting high-temperature REBCO superconductors using mechanical joints, the divertor is composed of a shower of molten tin jets, and the blanket is composed of molten salt FLiNaBe including Ti powers to increase hydrogen solubility. The main targets of the challenging option are early construction and easy maintenance of a large and three-dimensionally complicated helical structure, high thermal efficiency, and, in particular, realistic feasibility of the helical reactor.

  5. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  6. Experimental Investigation of Effect on Hydrate Formation in Spray Reactor

    Directory of Open Access Journals (Sweden)

    Jianzhong Zhao

    2015-01-01

    Full Text Available The effects of reaction condition on hydrate formation were conducted in spray reactor. The temperature, pressure, and gas volume of reaction on hydrate formation were measured in pure water and SDS solutions at different temperature and pressure with a high-pressure experimental rig for hydrate formation. The experimental data and result reveal that additives could improve the hydrate formation rate and gas storage capacity. Temperature and pressure can restrict the hydrate formation. Lower temperature and higher pressure can promote hydrate formation, but they can increase production cost. So these factors should be considered synthetically. The investigation will promote the advance of gas storage technology in hydrates.

  7. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H.; Enoeda, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

  8. Progress on source term evaluation of accidental events in the experimental fusion installation ITER

    Energy Technology Data Exchange (ETDEWEB)

    Virot, F.; Barrachin, M.; Vola, D.

    2015-10-15

    Highlights: • Progress of the IRSN R&D activities related to the safety assessment of the ITER installation. • Simulation of an accidental scenario with the ASTEC code: loss of coolant in port cell and in vacuum vessel. • Location and chemical speciation of beryllium dusts and tritium. - Abstract: The French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) in support to the French nuclear safety authority performs the safety analyses of the ITER experimental installation. We present the progress in the R&D activities related to a better evaluation of the source term in the event of an accident in this installation. These improvements are illustrated by an evaluation of the source term of a LOCA transient with the dedicated ASTEC code.

  9. Overview of the TIBER 2 (Thermal Ignition/Burn Experimental Reactor) design

    Science.gov (United States)

    Henning, C. D.; Logan, B. G.

    1987-10-01

    The TIBER 2 Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. While subsystems will be covered extensively in other reports, this overview will attempt to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER 2 can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition, 10 MW of ECR is added for disruption control and current drive profiling. The TIBER 2 design has been the US option in preparation for the International Thermonuclear Experimental Reactor (ITER). Other equivalent national designs are the NET in Europe, the FER in Japan and the OTR in the USSR. These designs will help set the basis for the new international design effort.

  10. Application of Iterative Robust Model-based Optimal Experimental Design for the Calibration of Biocatalytic Models

    DEFF Research Database (Denmark)

    Van Daele, Timothy; Gernaey, Krist V.; Ringborg, Rolf Hoffmeyer

    2017-01-01

    The aim of model calibration is to estimate unique parameter values from available experimental data, here applied to a biocatalytic process. The traditional approach of first gathering data followed by performing a model calibration is inefficient, since the information gathered during...... experimentation is not actively used to optimise the experimental design. By applying an iterative robust model-based optimal experimental design, the limited amount of data collected is used to design additional informative experiments. The algorithm is used here to calibrate the initial reaction rate of an ω......-transaminase catalysed reaction in a more accurate way. The parameter confidence region estimated from the Fisher Information Matrix is compared with the likelihood confidence region, which is a more accurate, but also a computationally more expensive method. As a result, an important deviation between both approaches...

  11. Reducing the effects of acoustic heterogeneity with an iterative reconstruction method from experimental data in microwave induced thermoacoustic tomography.

    Science.gov (United States)

    Wang, Jinguo; Zhao, Zhiqin; Song, Jian; Chen, Guoping; Nie, Zaiping; Liu, Qing-Huo

    2015-05-01

    An iterative reconstruction method has been previously reported by the authors of this paper. However, the iterative reconstruction method was demonstrated by solely using the numerical simulations. It is essential to apply the iterative reconstruction method to practice conditions. The objective of this work is to validate the capability of the iterative reconstruction method for reducing the effects of acoustic heterogeneity with the experimental data in microwave induced thermoacoustic tomography. Most existing reconstruction methods need to combine the ultrasonic measurement technology to quantitatively measure the velocity distribution of heterogeneity, which increases the system complexity. Different to existing reconstruction methods, the iterative reconstruction method combines time reversal mirror technique, fast marching method, and simultaneous algebraic reconstruction technique to iteratively estimate the velocity distribution of heterogeneous tissue by solely using the measured data. Then, the estimated velocity distribution is used subsequently to reconstruct the highly accurate image of microwave absorption distribution. Experiments that a target placed in an acoustic heterogeneous environment are performed to validate the iterative reconstruction method. By using the estimated velocity distribution, the target in an acoustic heterogeneous environment can be reconstructed with better shape and higher image contrast than targets that are reconstructed with a homogeneous velocity distribution. The distortions caused by the acoustic heterogeneity can be efficiently corrected by utilizing the velocity distribution estimated by the iterative reconstruction method. The advantage of the iterative reconstruction method over the existing correction methods is that it is successful in improving the quality of the image of microwave absorption distribution without increasing the system complexity.

  12. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  13. Application of iterative robust model-based optimal experimental design for the calibration of biocatalytic models.

    Science.gov (United States)

    Van Daele, Timothy; Gernaey, Krist V; Ringborg, Rolf H; Börner, Tim; Heintz, Søren; Van Hauwermeiren, Daan; Grey, Carl; Krühne, Ulrich; Adlercreutz, Patrick; Nopens, Ingmar

    2017-09-01

    The aim of model calibration is to estimate unique parameter values from available experimental data, here applied to a biocatalytic process. The traditional approach of first gathering data followed by performing a model calibration is inefficient, since the information gathered during experimentation is not actively used to optimize the experimental design. By applying an iterative robust model-based optimal experimental design, the limited amount of data collected is used to design additional informative experiments. The algorithm is used here to calibrate the initial reaction rate of an ω-transaminase catalyzed reaction in a more accurate way. The parameter confidence region estimated from the Fisher Information Matrix is compared with the likelihood confidence region, which is not only more accurate but also a computationally more expensive method. As a result, an important deviation between both approaches is found, confirming that linearization methods should be applied with care for nonlinear models. © 2017 American Institute of Chemical Engineers Biotechnol. Prog., 33:1278-1293, 2017. © 2017 American Institute of Chemical Engineers.

  14. ITER power electrical networks; Sistemas electricos de alimentacion a los consumidores del ITER

    Energy Technology Data Exchange (ETDEWEB)

    Sejas Portela, S.

    2011-07-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  15. Experimental and numerical study of the pressure drop for ITER blanket shield block

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Min-Su; Kim, Sawoong; Jung, Hun-Chea; Shim, Hee-Jin; Ahn, Hee-Jae

    2016-11-01

    Highlights: • The results of the experiment and the numerical analysis are compared. • The numerical analysis results are lower than the experimental results. • The margin of the pressure drop is suggested. - Abstract: The blanket shield block (SB) is located inside the ITER vacuum chamber, and the main function is to provide the thermal and nuclear shielding to the vacuum vessel and external components. The SB is foreseen to undergo a significant heat load which is a body load throughout the whole thickness of the SB under normal operation conditions. Therefore, the cooling configuration in SB should be designed very carefully based on the various experiences. The pressure drop in the cooling design is one of the most important factors to balance a water distribution of overall blanket cooling system. In order to verify the pressure drop characteristic and validate the design methodology of SB, experiment and numerical analysis are performed and compared their results. These results would be a benchmarking of the numerical results with experimental results to assess the gap between calculations and experiments.

  16. ITER driver blanket, European Community design

    Energy Technology Data Exchange (ETDEWEB)

    Simbolotti, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Zampaglione, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Ferrari, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Gallina, M. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Mazzone, G. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Nardi, C. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Petrizzi, L. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Rado, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Violante, V. (EURATOM-ENEA Association on Fusion Research, C.R.E., Frascati (Italy)); Daenner, W. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Lorenzetto, P. (NET Team, Max-Planck-Inst. fuer Plasmaphysik, Garching (Germany)); Gierszewski, P. (CFFTP, Mississauga, ON (Canada)); Gratt

    1993-07-01

    Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly afftect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. With the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper. (orig.)

  17. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H/sub 2/O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters.

  18. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  19. Experimental activity on the definition of acceptance criteria for the ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F. [CEA, IRFM, F-13108 St Paul Lez Durance (France)], E-mail: frederic.escourbiac@cea.fr; Constans, S. [SOM Ortec, Marseille (France); Vignal, N.; Cantone, V.; Richou, M.; Durocher, A. [CEA, IRFM, F-13108 St Paul Lez Durance (France); Riccardi, B. [Fusion For Energy, Barcelona (Spain); Bobin, I.; Jouvelot, J.L. [AREVA-NP, Le Creusot (France); Merola, M. [ITER Organization, Cadarache (France)

    2009-06-15

    Tens of thousands of armor/heat sink joints will be produced by the industry during the manufacturing of ITER divertor PFC, statistically, there is a probability that joints with defects be delivered. The purpose of this paper is to study the detection and evolution during operation of calibrated defects artificially implemented on samples, as an experimental basis for the definition of acceptance criteria for the bond armor/heat sink in the frame of industrial manufacturing conditions.It was found that current CFC monoblock design option was compatible with the heat loads specified at the lower part of the vertical target (up to 20 MW/m{sup 2}), including the presence of armor/heat sink defects (up to 50 deg. extension for a location at 0 deg. or 45 deg.) detectable with NDE techniques developed in Europe (US, SATIR). The current W monoblock design appeared suitable for the upper part of the vertical target with defects extension up to 50 deg. but is not adapted for heat flux of 20 MW/m{sup 2}. The studied W flat tile design proved to be compatible with fluxes of 5 MW/m{sup 2} but unable to sustain cycling fluxes of 10 MW/m{sup 2}.

  20. Introduction of Nuclear Instrumentations and Radiation Measurements in Experimental Fast Reactor 「JOYO」

    OpenAIRE

    大戸 敏弘; 鈴木 惣十

    1992-01-01

    This report introduces the nuclear instrumentation system and major R&D (research and development) activities using radiation measurement techniques in Experimental Fast Reactor "JOYO". In the introduction of the nuclear instrumentation system, following items are described; (1)system function (2)roles as a reactor plant equipment (3)specifications and charactelistics of neutron detectors, (4)construction and layout of the system. For reactor dosimetry at various irradiation tests and surveil...

  1. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  2. Review of accident analyses of RB experimental reactor

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2003-01-01

    Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

  3. A reverse flow catalytic membrane reactor for the production of syngas: an experimental study

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    In this paper experimental results are presented for a demonstration unit of a recently proposed novel integrated reactor concept (Smit et. al., 2005) for the partial oxidation of natural gas to syngas (POM), namely a Reverse Flow Catalytic Membrane Reactor (RFCMR). Natural gas has great potential

  4. Experimental and simulated dosimetry of the university of Utah TRIGA reactor

    Science.gov (United States)

    Marble, Benjamin James

    Simulated neutron and gamma transport enable the gamma dose to be estimated at the surface of the University of Utah TRIGA Reactor UUTR pool. These results are benchmarked against experimental results for model verification. This model is useful for future licensing and possible reactor power upgrades. MCNP5 was utilized for the UUTR simulation and comparison with thermoluminescent detectors TLDs.

  5. Transmission thermography for inspecting the busbar insulation layer in thermonuclear experimental reactor

    Science.gov (United States)

    Chen, Dapeng; Zhang, Guang; Zhang, Xiaolong; Zeng, Zhi

    2014-11-01

    In Thermonuclear Experimental Reactor, Superconducting Busbar is used for current transmission between magnet coils and current leads. The work temperature of the Busbar is about 4K because of liquid helium via inside. The large temperature grad from 300K to 4K could lead to the defects and damages occur on the insulation layer, which is made of glass fiber and polyimide and has a big different thermal expansion coefficient compared with the metal inner cylinder. This paper aims at developing an infrared transmission non-destructive evaluation (NDE) method for inspecting the insulation layer of Superconducting Busbar; theoretical model of transient heat conduction under a continuous inner heat source for cylindrical structure is described in the paper; a Busbar specimen which is designed with three delamination defects of different depths is heated inside by pouring hot water and monitored by an infrared detector located outside. Results demonstrate excellent detection performance for delamination defects in the insulation layer by using transmission thermography, all of the three defects of different depths can be visualized clearly in the thermal images, and the deeper defect has a better signal contrast, which is also shown in the temperature difference between defects and sound area vs. time curves. The results of light pulse thermography is also shown as a comparison, and it is found that the thermal images obtained by the transmission thermography has a much better signal contrast than that of the pulse thermography. In order to verify the experiments, finite element method is applied to simulate the heat conduction in the Busbar under the continuous inside heating, and it is found that the simulated temperature vs. time and simulated temperature difference vs. time curves are basically coincident with the experimental results. In addition, the possibility of in-service inspection for Busbar insulation layer in ITER item is discussed.

  6. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  7. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow

    2002-05-01

    The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV

  8. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    Energy Technology Data Exchange (ETDEWEB)

    Wuerz, H. [Kernforschungszentrum Karlsruhe, INR, Postfach 36 40, D-76021 Karlsruhe (Germany); Arkhipov, N.I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Bakhtin, V.P. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Konkashbaev, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Landman, I. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Safronov, V.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Toporkov, D.A. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation); Zhitlukhin, A.M. [Troitsk Institute for Innovation and Fusion Research, 142092 Troitsk (Russian Federation)

    1995-04-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.)).

  9. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Modeling of secondary emission processes in the negative ion based electrostatic accelerator of the International Thermonuclear Experimental Reactor

    Directory of Open Access Journals (Sweden)

    G. Fubiani

    2008-01-01

    Full Text Available The negative ion electrostatic accelerator for the neutral beam injector of the International Thermonuclear Experimental Reactor (ITER is designed to deliver a negative deuterium current of 40 A at 1 MeV. Inside the accelerator there are several types of interactions that may create secondary particles. The dominating process originates from the single and double stripping of the accelerated negative ion by collision with the residual molecular deuterium gas (≃29% losses. The resulting secondary particles (positive ions, neutrals, and electrons are accelerated and deflected by the electric and magnetic fields inside the accelerator and may induce more secondaries after a likely impact with the accelerator grids. This chain of reactions is responsible for a non-negligible heat load on the grids and must be understood in detail. In this paper, we will provide a comprehensive summary of the physics involved in the process of secondary emission in a typical ITER-like negative ion electrostatic accelerator together with a precise description of the numerical method and approximations involved. As an example, the multiaperture-multigrid accelerator concept will be discussed.

  11. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

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

  12. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    OpenAIRE

    Udiyani, P.M; Sri Kuntjoro

    2017-01-01

    Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimati...

  13. On the safety of ITER accelerators.

    Science.gov (United States)

    Li, Ge

    2013-01-01

    Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

  14. Experimental Investigation of Biogas Reforming in Gliding Arc Plasma Reactors

    Directory of Open Access Journals (Sweden)

    P. Thanompongchart

    2014-01-01

    Full Text Available Biogas is an important renewable energy source. Its utilization is restricted to vicinity of farm areas, unless pipeline networks or compression facilities are established. Alternatively, biogas may be upgraded into synthetic gas via reforming reaction. In this work, plasma assisted reforming of biogas was investigated. A laboratory gliding arc plasma setup was developed. Effects of CH4/CO2 ratio (1, 2.33, 9, feed flow rate (16.67–83.33 cm3/s, power input (100–600 W, number of reactor, and air addition (0–60% v/v on process performances in terms of yield, selectivity, conversion, and energy consumption were investigated. High power inputs and long reaction time from low flow rates, or use of two cascade reactors were found to promote dry reforming of biogas. High H2 and CO yields can be obtained at low energy consumption. Presence of air enabled partial oxidation reforming that produced higher CH4 conversion, compared to purely dry CO2 reforming process.

  15. Status of ITER task T213 collaborative irradiation screening experiment on Cu/SS joints in the Russian Federation SM-2-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Fabritsiev, S.A. [D.V. Efremov Inst., St. Petersburg (Russian Federation); Pokrovsky, A.S. [SRIAR, Dimitrovgrad (Russian Federation); Zinkle, S.J. [Oak Ridge National Lab., TN (United States)] [and others

    1996-04-01

    Specimen fabrication is underway for an irradiation screening experiment planned to start in January 1996 in the SM-2 reactor in Dimitrovgrad, Russia. The purpose of the experiment is to evaluate the effects of neutron irradiation at ITER-relevant temperatures on the bond integrity performance of Cu/SS and Be/Cu joints, as well as to further investigate the base metal properties of irradiated copper alloys. Specimens from each of the four ITER parties (U.S., EU, japan, and RF) will be irradiated to a dose of {approx}0.2 dpa at two different temperatures, 150 and 300{degrees}C. The specimens will consist of Cu/SS and Be/Cu joints in several different geometries, as well as a large number of specimens from the base materials. Fracture toughness data on base metal and Cu/SS bonded specimens will be obtained from specimens supplied by the U.S. Due to lack of material, the Be/Cu specimens supplied by the U.S will only be irradiated as TEM disks.

  16. Experiments in the experimental fast reactor VENUS-F: The FREYA project; Experimentos en el reactor rapido experimental VENUS-F: El proyecto FREYA

    Energy Technology Data Exchange (ETDEWEB)

    Villamarin, D.; Becares, V.; Cano, D.; Gonzalez, E.

    2011-07-01

    Due to the high flexibility of operation of the reactor VENUS-E, FREYA project has two main objectives. The first is the end of the study monitoring techniques reactivity and serve as validation of simulation codes. The second objective is to provide experimental support for design and licensing MYRRHA / FASTEE and TRF in collaboration with CDTy LEADER projects of the 7th Framework Programme of the EU.

  17. Experimental optimisation of a hypervapotron[reg ] concept for ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, Frederic E-mail: frederic.escourbiac@cea.fr; Schlosser, J.; Merola, M.; Bobin Vastra, I

    2003-09-01

    Historically developed for cooling of klystron electronic tubes, metallic hypervapotron[reg ] prototypes with different width were manufactured for high heat flux plasma facing components (PFC) applications. They were critical heat flux (CHF) tested on the European 200 kW electron beam facility (FE200), the 54 measured values have shown their good performances--up to 25-30 MW/m{sup 2} at low axial velocities (2-6 m/s)--interesting for ITER divertor dome and vertical target design. An important conclusion is that CHF decreases when the width increases.

  18. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    Science.gov (United States)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  19. Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

    2003-01-01

    In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...

  20. Experimental and theoretical studies of iterative methods for nonlinear, nonsymmetric systems arising in combustion

    Energy Technology Data Exchange (ETDEWEB)

    Hagstrom, T. [Univ. of New Mexico, Albuquerque, NM (United States); Radhakrishnan, K. [Sverdrup Technology, Brook Park, OH (United States)

    1994-12-31

    The authors report on some iterative methods which they have tested for use in combustion simulations. In particular, they have developed a code to solve zero Mach number reacting flow equations with complex reaction and diffusion physics. These equations have the form of a nonlinear parabolic system coupled with constraints. In semi-discrete form, one obtains DAE`s of index two or three depending on the number of spatial dimensions. The authors have implemented a fourth order (fully implicit) BDF method in time, coupled with a suite of fourth order explicit and implicit spatial difference approximations. Most codes they know of for simulating reacting flows use a splitting strategy to march in time. This results in a sequence of nonlinear systems to solve, each of which has a simpler structure than the one they are faced with. The rapid and robust solution of the coupled system is the essential requirement for the success of their approach. They have implemented and analyzed nonlinear generalizations of conjugate gradient-like methods for nonsymmetric systems, including CGS and the quasi-Newton based method of Eirola and Nevanlinna. They develop a general framework for the nonlinearization of linear methods in terms of the acceleration of fixed-point iterations, where the latter is assumed to include the {open_quote}preconditioning{open_quote}. Their preconditioning is a single step of a split method, using lower order spatial difference approximations as well as simplified (Fickian) approximations of the diffusion physics.

  1. On the hydraulic behaviour of ITER Shield Blocks #14 and #08. Computational analysis and comparison with experimental tests

    Energy Technology Data Exchange (ETDEWEB)

    Di Maio, P.A. [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy); Merola, M.; Mitteau, R.; Raffray, R. [ITER Organization, Route de Vinon sur Verdon, 13067 Saint Paul, Lez Durance (France); Vallone, E., E-mail: eug.vallone@gmail.com [Dipartimento di Energia, Ingegneria dell’Informazione e Modelli Matematici, Università di Palermo Viale delle Scienze, 90128, Palermo (Italy)

    2016-11-01

    Highlights: • A benchmarking activity has been carried out focusing the attention on the cooling circuits of ITER Shield Blocks #08 and #14. • A theoretical-computational fluid-dynamic approach based on the Finite Volume Method has been followed, adopting a commercial code. • Hydraulic characteristic functions and spatial distributions of coolant mass flow rate, velocity and pressure drop have been assessed. • Results obtained have allowed code benchmarking for Blanket modules and the numerical predictions have been found to be generally lower than but quite close to the experimental results (lower than 10%). - Abstract: As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are

  2. Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Germán, E-mail: german.perez.pichel@gmail.com; Mitteau, Raphaël; Eaton, Russell; Raffray, René

    2015-12-15

    Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

  3. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  4. Experimental characterization of slurry bubble-column reactor hydrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Shollenberger, K.A.; Torczynski, J.R.; Jackson, N.B.; O`Hern, T.J.

    1997-09-01

    Sandia`s program to develop, implement, and apply diagnostics for hydrodynamic characterization of slurry bubble column reactors (SBCRs) at industrially relevant conditions is discussed. Gas liquid flow experiments are performed on an industrial scale. Gamma densitometry tomography (GDT) is applied to measure radial variations in gas holdup at one axial location. Differential pressure (DP) measurements are used to calculate volume averaged gas holdups along the axis of the vessel. The holdups obtained from DP show negligible axial variation for water but significant variations for oil, suggesting that the air water flow is fully developed (minimal flow variations in the axial direction) but that the air oil flow is still developing at the GDT measurement location. The GDT and DP gas holdup results are in good agreement for the air water flow but not for the air oil flow. Strong flow variations in the axial direction may be impacting the accuracy of one or both of these techniques. DP measurements are also acquired at high sampling frequencies (250 Hz) and are interpreted using statistical analyses to determine the physical mechanism producing each frequency component in the flow. This approach did not yield the information needed to determine the flow regime in these experiments. As a first step toward three phase material distribution measurements, electrical impedance tomography (EIT) and GDT are applied to a liquid solid flow to measure solids holdup. Good agreement is observed between both techniques and known values.

  5. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  6. The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

    Science.gov (United States)

    Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

    2018-01-01

    The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

  7. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  8. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  9. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  10. The manufacture of enriched uranium fuel slugs for the Experimental Breeder Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Shuck, Author B.

    1953-04-20

    This report describes the specifications, materials and the sequence of operations used to found and fabricate 4 the first charge of enriched uranium fuel in the Experimental Breeder Reactor. The work was governed by the following principles: a. That the fuel be of correct composition, dimension and metallurgical condition for use in the reactor. b. That a maximum yield of finished fuel slugs from the quantity of uranium available for the program be achieved. c. That the residues be in a form which can be recovered by chemical or other means. d. That a detailed record be kept in such form that a complete history of each fuel slug be available.

  11. Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Broden, K. [Association EURATOM, Nykoeping (Sweden); Lindberg, M. [Association EURATOM, Nykoeping (Sweden); Nisan, S. [The NET Team, Garching (Germany); Rocco, P. [European Commission, Institute for Advanced Materials, Joint Research Centre, Ispra (Vatican City State, Holy See) (Italy); Zucchetti, M. [Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Taylor, N. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Forty, C. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    1997-04-01

    An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor and also for future fusion power station designs currently under investigation within the European programme on the safety and environmental assessment of fusion power, long-term. (orig.)

  12. Identification of new IκBα complexes by an iterative experimental and mathematical modeling approach.

    Directory of Open Access Journals (Sweden)

    Fabian Konrath

    2014-03-01

    Full Text Available The transcription factor nuclear factor kappa-B (NFκB is a key regulator of pro-inflammatory and pro-proliferative processes. Accordingly, uncontrolled NFκB activity may contribute to the development of severe diseases when the regulatory system is impaired. Since NFκB can be triggered by a huge variety of inflammatory, pro-and anti-apoptotic stimuli, its activation underlies a complex and tightly regulated signaling network that also includes multi-layered negative feedback mechanisms. Detailed understanding of this complex signaling network is mandatory to identify sensitive parameters that may serve as targets for therapeutic interventions. While many details about canonical and non-canonical NFκB activation have been investigated, less is known about cellular IκBα pools that may tune the cellular NFκB levels. IκBα has so far exclusively been described to exist in two different forms within the cell: stably bound to NFκB or, very transiently, as unbound protein. We created a detailed mathematical model to quantitatively capture and analyze the time-resolved network behavior. By iterative refinement with numerous biological experiments, we yielded a highly identifiable model with superior predictive power which led to the hypothesis of an NFκB-lacking IκBα complex that contains stabilizing IKK subunits. We provide evidence that other but canonical pathways exist that may affect the cellular IκBα status. This additional IκBα:IKKγ complex revealed may serve as storage for the inhibitor to antagonize undesired NFκB activation under physiological and pathophysiological conditions.

  13. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  14. On the possibility of experimentally confirming the hypothesis of reactor antineutrino passage into a sterile state

    Science.gov (United States)

    Serebrov, A. P.; Fomin, A. K.; Zinov'ev, V. G.; Loginov, Yu. E.; Onegin, M. S.; Gagarskiy, A. M.; Petrov, G. A.; Solovei, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martem'yanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Svyatkin, M. N.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Khramkov, N. S.; Rykalin, V. I.

    2013-07-01

    The "Neutrino-4" experiment for the 100-MW SM-3 reactor has been developed with the aim of testing the reactor antineutrino anomaly at Petersburg Nuclear Physics Institute. The advantages of this reactor for studying the antineutrino anomaly are (i) a low background level and (ii) small dimensions (35 × 42 × 42 cm) of the active zone. Operation of a position-sensitive antineutrino detector comprising five working sections and moving so as to cover a region of distances within 6-13 m from the active zone has been simulated by the Monte-Carlo method. The range of experimental sensitivity with respect to the oscillation parameters Δ m 2 and sin22θ is determined, which will make it possible to confirm the hypothesis of antineutrino oscillations into a sterile state.

  15. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    Energy Technology Data Exchange (ETDEWEB)

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  16. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  17. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul Lam; Dimitri Gidaspow

    2000-09-01

    The objective if this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The computed time averaged particle velocities and concentrations agree with PIV measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. This phase of the work was presented at the Chemical Reaction Engineering VIII: Computational Fluid Dynamics, August 6-11, 2000 in Quebec City, Canada. To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV technique. The results together with simulations will be presented at the annual meeting of AIChE in November 2000.

  18. Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

    KAUST Repository

    Wang, Jui-Yang

    2017-06-01

    A jet-stirred reactor was designed and constructed in the Clean Combustion Research Center (CCRC) at King Abdullah University of Science and Technology (KAUST); was validated with n-heptane, iso-octane oxidation and cyclohexene pyrolysis. Different configurations of the setup have been tested to achieve good agreement with results from the literature. Test results of the reactor indicated that installation of a pumping system at the downstream side in the experimental apparatus was necessary to avoid the reoccurrence of reactions in the sampling probe. Experiments in ethyl levulinate oxidation were conducted in the reactor under several equivalence ratios, from 600 to 1000 K, 1 bar and 2 s residence time. Oxygenated species detected included methyl vinyl ketone, levulinic acid and ethyl acrylate. Ethylene, methane, carbon monoxide, hydrogen, oxygen and carbon dioxide were further quantified with a gas chromatography, coupled with a flame ionization detector and a thermal conductivity detector. The ethyl levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental data, some discrepancies were noted; predictions of ethylene production were not well matched. The kinetic model was improved by updating several classes of reactions: unimolecular decomposition, H-abstraction, C-C and C-O beta-scissions of fuel radicals. The updated model was then compared again with experimental results and good agreement was achieved, proving that the concerted eliminated reaction is crucial for the kinetic mechanism formulation of ethyl levulinate. In addition, primary reaction pathways and sensitivity analysis were performed to describe the role of molecular structure in combustion (800 and 1000 K for ethyl levulinate oxidation in the jet-stirred reactor).

  19. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    Science.gov (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  20. Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

    Science.gov (United States)

    Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

    2017-08-01

    The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

  1. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    OpenAIRE

    Sayer C.; Palma M.; Giudici R.

    2002-01-01

    A new reactor, the pulsed sieve plate column (PSPC), was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA) emulsion polymerization reactions performed in this PSPC. Therefore, both experimental ...

  2. Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Utgikar, Vivek [Univ. of Idaho, Moscow, ID (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Christensen, Richard [The Ohio State Univ., Columbus, OH (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-29

    The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate the models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.

  3. Development of a membrane-assisted fluidized bed reactor - 2 - Experimental demonstration and modeling for the partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Laverman, J.A.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    A small laboratory-scale membrane-assisted fluidized bed reactor (MAFBR) was constructed in order to experimentally demonstrate the reactor concept for the partial oxidation of methanol to formaldehyde. Methanol conversion and product selectivities were measured at various overall fluidization

  4. Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2011-01-01

    Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

  5. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-10-26

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended.

  6. Prospects for Tokamak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  7. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  8. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)

    2012-11-01

    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to {proportional_to}1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  9. Experimental measurement and CFD simulation on the hydrodynamics of an internal-loop airlift reactor

    Directory of Open Access Journals (Sweden)

    Liew Shi Yan

    2017-01-01

    Full Text Available This paper concerns with the experimental measurement and computational fluid dynamics simulation on local hydrodynamics of a gas-liquid internal-loop airlift reactor. The aim of this work is to study the sensitivity of the drag models and the significance of considering the lift force on the predictive accuracy of the simulation. The experimental analysis was carried out using laser Doppler anemometry at three different heights (i.e. Y = 0.20 m, 0.30 m and 0.38 m across the riser and downcomerat volumetric flow rate of 0.30 m3/h to provide validation for the simulation results. A transient three-dimensional gasliquid internal-loop airlift reactor was carried out using FLUENT 16.2 by implementing the two-fluid model approach. The Eulerian-Eulerian multiphase and standard κ-ε dispersed turbulence model wereemployed in this study. Results suggest that the spherical drag model performed poorly and that the drag model governed by Rayleigh-Taylor shows promising accuracy in the prediction of overall mean axial liquid velocity. On the other hand, the consideration of lift model shows slightly improvement in accuracy. These findings may serve as a guidance for future scale-up and design of airlift reactor studies

  10. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  11. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  12. Spirit and prospects of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E.P. [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

    2002-10-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  13. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  14. Experimental conditions for determination of the neutrino mass hierarchy with reactor antineutrinos

    Directory of Open Access Journals (Sweden)

    Myoung Youl Pac

    2016-01-01

    Full Text Available This article reports the optimized experimental requirements to determine neutrino mass hierarchy using electron antineutrinos (ν¯e generated in a nuclear reactor. The features of the neutrino mass hierarchy can be extracted from the |Δm312| and |Δm322| oscillations by applying the Fourier sine and cosine transforms to the L/E spectrum. To determine the neutrino mass hierarchy above 90% probability, the requirements on the energy resolution as a function of the baseline are studied at sin2⁡2θ13=0.1. If the energy resolution of the neutrino detector is less than 0.04/Eν and the determination probability obtained from Bayes' theorem is above 90%, the detector needs to be located around 48–53 km from the reactor(s to measure the energy spectrum of ν¯e. These results will be helpful for setting up an experiment to determine the neutrino mass hierarchy, which is an important problem in neutrino physics.

  15. Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

    2017-09-01

    The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

  16. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    Directory of Open Access Journals (Sweden)

    P.M. Udiyani

    2017-02-01

    Full Text Available Experimental power reactor (RDE which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated using ORIGEN-2 whose library has been modified for HTGR type, and the assumptions of defects of the TRISO fuel and release fraction for each compartment of RDE safety system used a reference parameter. The results showed that the important source terms of RDE are group of noble gases (Kr and Xe, halogen (I, Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTGR source terms with the same power. Keywords: routine discharge, radionuclide, source term, RDE, HTGR

  17. Model-Free Primitive-Based Iterative Learning Control Approach to Trajectory Tracking of MIMO Systems With Experimental Validation.

    Science.gov (United States)

    Radac, Mircea-Bogdan; Precup, Radu-Emil; Petriu, Emil M

    2015-11-01

    This paper proposes a novel model-free trajectory tracking of multiple-input multiple-output (MIMO) systems by the combination of iterative learning control (ILC) and primitives. The optimal trajectory tracking solution is obtained in terms of previously learned solutions to simple tasks called primitives. The library of primitives that are stored in memory consists of pairs of reference input/controlled output signals. The reference input primitives are optimized in a model-free ILC framework without using knowledge of the controlled process. The guaranteed convergence of the learning scheme is built upon a model-free virtual reference feedback tuning design of the feedback decoupling controller. Each new complex trajectory to be tracked is decomposed into the output primitives regarded as basis functions. The optimal reference input for the control system to track the desired trajectory is next recomposed from the reference input primitives. This is advantageous because the optimal reference input is computed straightforward without the need to learn from repeated executions of the tracking task. In addition, the optimization problem specific to trajectory tracking of square MIMO systems is decomposed in a set of optimization problems assigned to each separate single-input single-output control channel that ensures a convenient model-free decoupling. The new model-free primitive-based ILC approach is capable of planning, reasoning, and learning. A case study dealing with the model-free control tuning for a nonlinear aerodynamic system is included to validate the new approach. The experimental results are given.

  18. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  19. Physics design of a 100 keV acceleration grid system for the diagnostic neutral beam for international tokamak experimental reactor.

    Science.gov (United States)

    Singh, M J; De Esch, H P L

    2010-01-01

    This paper describes the physics design of a 100 keV, 60 A H(-) accelerator for the diagnostic neutral beam (DNB) for international tokamak experimental reactor (ITER). The accelerator is a three grid system comprising of 1280 apertures, grouped in 16 groups with 80 apertures per beam group. Several computer codes have been used to optimize the design which follows the same philosophy as the ITER Design Description Document (DDD) 5.3 and the 1 MeV heating and current drive beam line [R. Hemsworth, H. Decamps, J. Graceffa, B. Schunke, M. Tanaka, M. Dremel, A. Tanga, H. P. L. De Esch, F. Geli, J. Milnes, T. Inoue, D. Marcuzzi, P. Sonato, and P. Zaccaria, Nucl. Fusion 49, 045006 (2009)]. The aperture shapes, intergrid distances, and the extractor voltage have been optimized to minimize the beamlet divergence. To suppress the acceleration of coextracted electrons, permanent magnets have been incorporated in the extraction grid, downstream of the cooling water channels. The electron power loads on the extractor and the grounded grids have been calculated assuming 1 coextracted electron per ion. The beamlet divergence is calculated to be 4 mrad. At present the design for the filter field of the RF based ion sources for ITER is not fixed, therefore a few configurations of the same have been considered. Their effect on the transmission of the electrons and beams through the accelerator has been studied. The OPERA-3D code has been used to estimate the aperture offset steering constant of the grounded grid and the extraction grid, the space charge interaction between the beamlets and the kerb design required to compensate for this interaction. All beamlets in the DNB must be focused to a single point in the duct, 20.665 m from the grounded grid, and the required geometrical aimings and aperture offsets have been calculated.

  20. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.; Silcox, Geoffrey D.

    2009-06-15

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  1. Modeling and experimental validation of hydrodynamics in an ultrasonic batch reactor.

    Science.gov (United States)

    Ajmal, M; Rusli, S; Fieg, G

    2016-01-01

    Simulation of hydrodynamics in ultrasonic batch reactor containing immobilized enzymes as catalyst is done. A transducer with variable power and constant frequency (24 kHz) is taken as source of ultrasound (US). Simulation comprises two steps. In first step, acoustic pressure field is simulated and in second step effect of this field on particle trajectories is simulated. Simulation results are compared with experimentally determined particle trajectories using PIV Lab (particle image velocimetry). Effect of varying ultrasonic power, positioning and number of ultrasonic sources on particle trajectories is studied. It is observed that catalyst particles tend to orientate according to pattern of acoustic pressure field. An increase in ultrasonic power increases particle velocity and also brings more particles into motion. Simulation results are found to be in agreement with experimentally determined data. Copyright © 2015 Elsevier B.V. All rights reserved.

  2. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2002-01-01

    Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.

  3. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

    The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

  4. CFD and experimental investigation of sloshing parameters for the safety assessment of HLM reactors

    Energy Technology Data Exchange (ETDEWEB)

    Myrillas, Konstantinos, E-mail: myrillas@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe, E-mail: philippe.planquart@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Simonini, Alessia, E-mail: Simonini@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Buchlin, Jean-Marie, E-mail: buchlin@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Schyns, Marc, E-mail: mschyns@SCKCEN.BE [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    Highlights: • Comparison of sloshing behavior in cylindrical tank using mercury and water. • Flow visualization of liquid sloshing in resonance case. • CFD simulations of sloshing with OpenFOAM, using the VOF method. • Qualitative and quantitative comparison of experimental and numerical results. • Evaluation of sloshing forces on the tank walls from numerical simulations. - Abstract: For the safety assessment of Heavy Liquid Metal nuclear reactors under seismic excitation, sloshing phenomena can be of great concern. The earthquake motions are transferred to the liquid coolant which oscillates inside the vessel, exerting additional forces on the walls and internal structures. The present study examines the case of MYRRHA, a multi-purpose experimental reactor with LBE as coolant, developed by SCK·CEN. The sloshing behavior of liquid metals is studied through a comparison between mercury and water in a cylindrical tank. Experimental investigation of sloshing is carried out using optical techniques with the shaking table facility SHAKESPEARE at the von Karman Institute. Emphasis is given on the resonance case, where maximum forces occur on the tank walls. The experimental cases are reproduced numerically with the CFD software OpenFOAM, using the VOF method to track the liquid interface. The non-linear nature of sloshing is observed through visualization, where swirling is shown in the resonance case. The complex behavior is well reproduced by the CFD simulations, providing good qualitative validation of the numerical tools. A quantitative comparison of the maximum liquid elevation inside the tank shows higher values for the liquid metal than for water. Some discrepancies are revealed in CFD results and the differences are quantified. From simulations it is verified that the forces scale with the density ratio, following similar evolution in time. Overall, water is demonstrated to be a valid option as a working liquid in order to evaluate the sloshing

  5. Experimental study and modeling of a high-temperature solar chemical reactor for hydrogen production from methane cracking

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, Stephane; Flamant, Gilles [Processes, Materials, and Solar Energy Laboratory, CNRS (PROMES-CNRS, UPR 8521), 7 Rue du Four Solaire, 66120 Odeillo Font-Romeu (France)

    2007-07-15

    A high-temperature fluid-wall solar reactor was developed for the production of hydrogen from methane cracking. This laboratory-scale reactor features a graphite tubular cavity directly heated by concentrated solar energy, in which the reactive flowing gas dissociates to form hydrogen and carbon black. The solar reactor characterization was achieved with: (a) a thorough experimental study on the reactor performance versus operating conditions and (b) solar reactor modeling. The results showed that the conversion of CH{sub 4} and yield of H{sub 2} can exceed 97% and 90%, respectively, and these depend strongly on temperature and on fluid-wall heat transfer and reaction surface area. In addition to the experimental study, a 2D computational model coupling transport phenomena was developed to predict the mapping of reactor temperature and of species concentration, and the reaction extent at the outlet. The model was validated and kinetics of methane decomposition were identified from simulations and comparison to experimental results. (author)

  6. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

  7. 14MeV neutron irradiation experiment on window materials for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Fuminobu; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iida, Toshiyuki

    1997-06-01

    Data on wavelength spectra of photons emitted from window material during neutron and gamma-ray irradiation has been required for design of next D-T burning fusion reactor such as ITER. Thus, a photon measurement system has been developed to analyze wavelength spectra of photons emitted from the optical window materials during 14MeV-neutron irradiation, and the system consisted of a sample holder, a radiation-resistant optical fiber, a photon counting analyzer and other electronic devices. The irradiation experiments for synthesized sapphire, high-purity silica glass and synthesized quartz were performed using a fusion neutron source FNS. As for all the sample, number of photon emission was proportional to the 14MeV-neutron flux in the range of 10{sup 6}-10{sup 11}n/cm{sup 2}/sec. The photon emission efficiency of F-center luminescence of the sapphire was 2200 {+-} 700photons/MeV, while the efficiency of F{sup +}-center luminescence was two order less than that of F-center. The wavelength spectra of the high-purity silica glass had a large peak around 450nm, which was concerned with decay of self-trapped excitons in oxygen vacancies. Its photon emission efficiency for 14MeV-neutrons has been found to be about 5 {+-} 3photons/MeV in visible range, while that for gamma-rays to be about 135 {+-} 50photons/MeV. The spectrum of photons emitted from the quartz had two large peaks around not only 450nm but also 650nm, and the photon emission efficiency in the wavelength range of 350-750nm was 14 {+-} 4photons/MeV. (author)

  8. Experimental validation of a lap-type joint AC loss model with an ITER correction coil conductor joint

    NARCIS (Netherlands)

    Rolando, G.; Hansheng, F.; Hongwei, L.; Lin, W.; Wu, W.; Foussat, A.; Ilyin, Y.; Libeyre, P.; Nijhuis, Arend

    2014-01-01

    The ITER correction coils (CC) system features shaking hands lap-type joints to interface the terminations of the conductors. The feasibility of operating plasma scenarios depends on the ability of the magnets to retain sufficient temperature and current margins. In this respect, the joints

  9. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    Energy Technology Data Exchange (ETDEWEB)

    Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L. (Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering); Kisner, R.A. (Oak Ridge National Lab., TN (United States))

    1991-08-01

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs.

  10. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    coolants. The purpose of the high temperature helium loop (HTHL) is to simulate technical and chemical conditions of VHTR's coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes also the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.

  11. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.

  12. A large scale fullerenes synthesis solar reactor modelling and first experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, T.; Flamand, G.; Robert, J.F.; Rivoire, B.; Olalde, G.; Alvarez, L. [Centre National de la Recherche Scientifique (CNRS-IMP), 66 - Font-Romeu (France); Laplaze, D. [Universite de Montpellier, GDPC, 34 (France)

    1999-03-01

    After the promising results obtained with a 2 kW solar furnace for fullerenes and nano-tubes synthesis, a large scale production project using the 1 MW Odeillo solar furnace started in 1997. This paper presents the first experimental results obtained with a concept-validation vessel and the comparison with a numerical simulation of the target thermal behavior. It is shown that a 6 mm i.d. graphite rod heated by a 500 W/cm{sup 2} incident solar flux density (I{sub s}) reaches a front temperature of 2800 K, in agreement with the thermal model. On this basis, accurate prediction of maximum working temperature of the 1 MW reactor is proposed: 3400 K for I{sub s} = 900 W/cm{sup 2}. (authors)

  13. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burns, Douglas Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kammerer, Annie [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

  14. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  15. Nuclear aspects of diagnostics in RTO/RC ITER

    Energy Technology Data Exchange (ETDEWEB)

    Walker, C.I. E-mail: walkerc@ipp-garching.mpg.de; Yamamoto, S.; Costley, A.; Kock, L. de; Ebisawa, K.; Janeschitz, G.; Khripunov, V.; Martin, E.; Vayakis, G

    2000-11-01

    ITER (international thermonuclear experimental reactor) will be the first fusion device where the design of the plasma diagnostic systems will make extensive use of the materials and techniques developed in the nuclear technology field. The designs have to satisfy stringent requirements for tritium confinement, nuclear shielding and vacuum integrity. This paper introduces the requirements for diagnostics in the ITER long pulse, burning plasma environment, and addresses the impact of the reactor environment on the diagnostics and ancillary equipment. These systems necessarily require access to the plasma or first wall, which generally conflicts with the requirements of the basic machine. Holes are required through the first wall, primary shielding, containment boundaries and biological shielding. Components have a limited life and require maintenance. This paper describes the effect of the radiation environment on diagnostic design at different locations. Ex-vessel and in-vessel remote handling, hot cell refurbishment and tritium confinement are also described.

  16. Nuclear performance of the D-D phase of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, V. E-mail: khripuv@itereu.de

    2000-11-01

    Deuterium plasma will be used during the early phase of the International Thermonuclear Experimental Reactor (ITER) operation. For a typical reduced cost ITER the D-D-neutron yield {approx}3.5x10{sup 18} n s{sup -1} will be accompanied by a considerable tritium production and a D-T (14.1-MeV)-neutron yield of {approx}0.7-3.5x10{sup 18} n s{sup -1}. This paper describes the nuclear performance for the two-component neutron source experienced in the proposed advanced H-mode D-D regimes. The calculated nuclear energy release and residual activity estimates can be important from the radiation safety, maintenance and staged ITER construction standpoints.

  17. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  18. 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).

  19. Analysis of the optimal fuel composition for the Indonesian experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.), Ibaraki (Japan); Sembiring, Tagor Malem [National Nuclear Energy Agency of Indonesia, Banten (Indonesia). Center for Nuclear Reactor Technology and Safety; Arbie, Bakri; Subki, Iyos [PT MOTAB Technology, Jakarta Barat (Indonesia)

    2017-03-15

    The optimal fuel composition of the 10 MWth Experimental Power Reactor (RDE), to be built by the Indonesian National Nuclear Energy Agency (BATAN), is a very important design parameter since it will directly affect the fuel cost, new and spent fuel storage capacity, and other back-end environmental burden. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO{sub 2} TRISO fuel under multipass or once-through-then-out fueling scheme. A scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment is conducted. All burnup, criticality calculations and core equilibrium search are carried out by using BATAN-MPASS, a general in-core fuel management code for pebble bed HTGRs, featured with many automatic equilibrium searching options as well as thermal-hydraulic calculation capability. The RDE User Requirement Document issued by BATAN is used to derive the main core design parameters and constraints. The scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4 g to 10 g/pebble. Fissile loading per unit energy generated (kg/GWd) is taken as the objective function for the present scoping study. The analysis results show that the optimal HM loading is around 8 g/pebble. Under the constraint of 80 GWd/t fuel discharge burnup imposed by the technical specification, the uranium enrichment for the optimal HM loading is approximately 13 w/o.

  20. Thermo-kinetic instabilities in model reactors. Examples in experimental tests

    Science.gov (United States)

    Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele

    2017-11-01

    The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and

  1. Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi (JAERI, Ibaraki (Japan))

    1993-06-01

    If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.

  2. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  3. Experimental and numerical studies of microwave-plasma interaction in a MWPECVD reactor

    Directory of Open Access Journals (Sweden)

    A. Massaro

    2016-12-01

    Full Text Available This work deals with and proposes a simple and compact diagnostic method able to characterize the interaction between microwave and plasma without the necessity of using an external diagnostic tool. The interaction between 2.45 GHz microwave and plasma, in a typical ASTeX-type reactor, is investigated from experimental and numerical view points. The experiments are performed by considering plasmas of three different gas mixtures: H2, CH4-H2 and CH4-H2-N2. The two latter are used to deposit synthetic undoped and n-doped diamond films. The experimental setup equipped with a matching network enables the measurements of very low reflected power. The reflected powers show ripples due to the mismatching between wave and plasma impedance. Specifically, the three types of plasma exhibit reflected power values related to the variation of electron-neutral collision frequency among the species by changing the gas mixture. The different gas mixtures studied are also useful to test the sensitivity of the reflected power measurements to the change of plasma composition. By means of a numerical model, only the interaction of microwave and H2 plasma is examined allowing the estimation of plasma and matching network impedances and of reflected power that is found about eighteen times higher than that measured.

  4. Iterant Algebra

    Directory of Open Access Journals (Sweden)

    Louis H. Kauffman

    2017-07-01

    Full Text Available We give an exposition of iterant algebra, a generalization of matrix algebra that is motivated by the structure of measurement for discrete processes. We show how Clifford algebras and matrix algebras arise naturally from iterants, and we then use this point of view to discuss the Schrödinger and Dirac equations, Majorana Fermions, representations of the braid group and the framed braids in relation to the structure of the Standard Model for physics.

  5. Study of process of water disinfection it saw energy solar using an experimental reactor; Estudo do proceso de desinfeccao de agua via energia solar utilizando um reator experimental

    Energy Technology Data Exchange (ETDEWEB)

    Batista, C. H.; Prado, L. R.; Lima, A. S.; Egues, S. M. S.; Araujo, P. M. M.

    2008-07-01

    In this work, was conducted an experimental study of the efficiency of a solar reactor in the disinfection of drinking water using photolysis (UV) and heterogeneous photo catalysis (TiO{sub 2}/UV). The experiments were conducted in batch mode, evaluating the effects of reactor inclination and the presence of a solar concentrator. The results indicated that the employed system was capable to promote the complete disinfection in 150 min using only the photo thermic effect, and in 120 min with the addition of immobilized TiO{sub 2} and the solar concentrator. (Author)

  6. Estudio del comportamiento de reactores discontinuos y semicontinuos: modelización y comprobación experimental

    OpenAIRE

    Grau Vilalta, Ma. Dolors

    1999-01-01

    L'objectiu primordial d'aquest treball és la comparació entre el funcionament d'un reactor discontinu i un de semicontinu. Per això es porta a terme la modelització matemàtica d'ambdós, utilitzant programes propis emprant el llenguatge Fortran 77, a més del simulador ISIM i el software MATLAB. La validació dels models matemàtics s'efectua, en primer lloc, a partir de dades de la bibliografia. A partir d'aquí, es realitzen proves experimentals en una planta piloto amb un reactor encamisat de v...

  7. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  8. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  9. DC performance and AC loss of cable-in-conduit conductors for International Thermonuclear Experimental Reactor

    National Research Council Canada - National Science Library

    Zhou, Wei; Fang, Xin-Yu; Fang, Jin; Liu, Bo

    2016-01-01

    ...) cable-in-conduit conductors(CICCs);however,the calculation of AC loss of ITER CICCs is a cumbersome task due to the complicated geometry of the multistage cables and the extreme operating conditions...

  10. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  11. Experimental and statistical investigation of thermally induced failure in reactor fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Lunsford, J.L.; Imprescia, R.J.; Bowman, A.L.; Radosevich, C.E.

    1980-10-01

    An incomplete experimental study into the failure statistics of fuel particle for the high-temperature gas-cooled reactor (HTGR) is described. Fuel particles failure was induced by thermal ramping from room temperature to temperatures in the vicinity of 2273/sup 0/K to 2773/sup 0/K in 2 to 30 h and detected by the appearance of /sup 85/Kr in the helium carrier gas used to sweep the furnace. The concentration of krypton, a beta emitter, was detected by measuring the current that resulted when the helium sweep gas was passed through an ionization chamber. TRISO fuel particles gave a krypton concentration profile as a function of time that built up in several minutes and decayed in a fraction of an hour. This profile, which was temperature independent, was similar to the impulse response of the ionization chamber, suggesting that the TRISO particles failed instantaneously and completely. BISO fuel particles gave a krypton concentration profile as a function of time that built up in a fraction of an hour and decayed in a fraction of a day. This profile was strongly temperature dependent, suggesting that krypton release was diffusion controlled, i.e., that the krypton was diffusing through a sound coat, or that the BISO coating failed but that the krypton was unable to escape the kernel without diffusion, or that a combination of pre- and postfailure diffusion accompanied partial or complete failure.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-07-15

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

  13. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  14. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    Science.gov (United States)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

  15. Design study of toroidal magnets for tokamak experimental power reactors. [NbTi alloys

    Energy Technology Data Exchange (ETDEWEB)

    Stekly, Z.J.J.; Lucas, E.J. (eds.)

    1976-12-01

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed.

  16. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  17. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    Directory of Open Access Journals (Sweden)

    Mark Daniel G. de Luna

    2017-05-01

    Full Text Available This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD wastewater. The effect of inorganic ions CO32−, NH4+ and Al3+ on sulfate conversion was also investigated. Experimental results indicated that seed dosage, initial sulfate concentration and molar ratio of calcium to sulfate are all significant parameters in the sulfate removal by fluidized-bed crystallization. The optimal conditions of 4 g seed L−1, 119.7 mM of initial sulfate concentration and [Ca2+]/[SO42−] molar ratio of 1.48 resulted in sulfate conversion of 82% and sulfate removal of 67%. Conversion and removal of sulfate in the simulated AMD wastewater were 79 and 63%, respectively. When ammonium or aluminum was added to the synthetic sulfate wastewater, significant conversion of sulfate was achieved.

  18. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn/sub 3/ conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended.

  19. A novel reverse flow reactor coupling endothermic and exothermic reactions: an experimental study

    NARCIS (Netherlands)

    van Sint Annaland, M.; Nijssen, R.C.

    2002-01-01

    A new reactor concept is studied for highly endothermic heterogeneously catalysed gas phase reactions at high temperatures with rapid but reversible catalyst deactivation. The reactor concept aims to achieve an indirect coupling of energy necessary for endothermic reactions and energy released by

  20. Experimental and numerical stability investigations on natural circulation boiling water reactors

    CERN Document Server

    Marcel, CP

    2007-01-01

    In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs

  1. Plasma cleaning of ITER first mirrors

    Science.gov (United States)

    Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

    2017-12-01

    Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

  2. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  3. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  4. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  5. Hydrocarbon pyrolysis reactor experimentation and modeling for the production of solar absorbing carbon nanoparticles

    Science.gov (United States)

    Frederickson, Lee Thomas

    Much of combustion research focuses on reducing soot particulates in emissions. However, current research at San Diego State University (SDSU) Combustion and Solar Energy Laboratory (CSEL) is underway to develop a high temperature solar receiver which will utilize carbon nanoparticles as a solar absorption medium. To produce carbon nanoparticles for the small particle heat exchange receiver (SPHER), a lab-scale carbon particle generator (CPG) has been built and tested. The CPG is a heated ceramic tube reactor with a set point wall temperature of 1100-1300°C operating at 5-6 bar pressure. Natural gas and nitrogen are fed to the CPG where natural gas undergoes pyrolysis resulting in carbon particles. The gas-particle mixture is met downstream with dilution air and sent to the lab scale solar receiver. To predict soot yield and general trends in CPG performance, a model has been setup in Reaction Design CHEMKIN-PRO software. One of the primary goals of this research is to accurately measure particle properties. Mean particle diameter, size distribution, and index of refraction are calculated using Scanning Electron Microscopy (SEM) and a Diesel Particulate Scatterometer (DPS). Filter samples taken during experimentation are analyzed to obtain a particle size distribution with SEM images processed in ImageJ software. These results are compared with the DPS, which calculates the particle size distribution and the index of refraction from light scattering using Mie theory. For testing with the lab scale receiver, a particle diameter range of 200-500 nm is desired. Test conditions are varied to understand effects of operating parameters on particle size and the ability to obtain the size range. Analysis of particle loading is the other important metric for this research. Particle loading is measured downstream of the CPG outlet and dilution air mixing point. The air-particle mixture flows through an extinction tube where opacity of the mixture is measured with a 532 nm

  6. Parametric experimental tests of steam gasification of pine wood in a fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    L. Vecchione

    2013-09-01

    Full Text Available Among Renewable Energy Sources (RES, biomass represent one of the most common and suitable solution in order to contribute to the global energy supply and to reduce greenhouse gases (GHG emissions. The disposal of some residual biomass, as pruning from pine trees, represent a problem for agricultural and agro-industrial sectors. But if the residual biomass are used for energy production can become a resource. The most suitable energy conversion technology for the above-mentioned biomass is gasification process because the high C/N ratio and the low moisture content, obtained from the analysis. In this work a small-pilot bubbling-bed gasification plant has been designed, constructed and used in order to obtain, from the pine trees pruning, a syngas with low tar and char contents and high hydrogen content. The activities showed here are part of the activities carried out in the European 7FP UNIfHY project. In particular the aim of this work is to develop experimental test on a bench scale steam blown fluidized bed biomass gasifier. These tests will be utilized in future works for the simulations of a pilot scale steam fluidized bed gasifier (100 kWth fed with different biomass feedstock. The results of the tests include produced gas and tar composition as well gas, tar and char yield. Tests on a bench scale reactor (8 cm I.D. were carried out varying steam to biomass ratio from 0.5, 0.7 and 1 to 830°C.

  7. Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

    Science.gov (United States)

    Opanasenko, A. N.; Sorokin, A. P.; Zaryugin, D. G.; Trufanov, A. A.

    2017-05-01

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.

  8. Experimental evidences of95 mTc production in a nuclear reactor.

    Science.gov (United States)

    Cohen, I M; Robles, A; Mendoza, P; Airas, R M; Montoya, E H

    2018-02-02

    95 m Tc has been identified as by-product in some solutions of 99 m Tc obtained by irradiation of molybdenum trioxide in a reactor neutron flux. The characterization was carried out using both measurements by gamma spectrometry and half-life determination. The possible ways that lead to the 95 m Tc production in a nuclear reactor are discussed. Copyright © 2018. Published by Elsevier Ltd.

  9. Experimental evaluation of gamma fluence-rate predictions from Argon-41 releases to the atmosphere over a nuclear research reactor site

    DEFF Research Database (Denmark)

    Rojas-Palma, C.; Aage, H.K.; Astrup, P.

    2004-01-01

    An experimental study of radionuclide dispersion in the atmosphere has been conducted at the BR1 research reactor in Mol, Belgium. Artificially generated aerosols ('white smoke') were mixed with the routine releases of Ar-41 in the reactor's 60-m tall venting stack. The detailed plume geometry...

  10. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  11. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  12. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  13. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  14. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  15. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  16. Experimental investigation into fast pyrolysis of biomass using an entrained flow reactor

    Science.gov (United States)

    Bohn, M.; Benham, C.

    1981-02-01

    Pyrolysis experiments were performed with steam as a carrier gas and two different feedstocks - wheat straw and powdered material derived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 7000 to 1400 C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock.

  17. Evaluation of activation detectors for the SPHINX project at the LR-0 experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, Zdena; Viererbl, Ladislav [Research Center Rez Ltd (Czech Republic); Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech [Nuclear Power and Safety Division, Nuclear Research Institute Rez plc (Czech Republic)

    2008-07-01

    This article summarizes the measurements of neutron fluence distributions carried out at the LR-0 research reactor (Czech Republic) in the frame of the SPHINX project. The influence of fluoride-salts or graphite filling in the SR-0 modules on neutron spectrum was studied using activation detectors. The activation detectors (Mn, Ni, In and Au) were evaluated to determine the changes in neutron field. The In and Au detectors were also irradiated with a cadmium cover. Five different configurations of reactor core (EROS) were realized. (authors)

  18. Experimental evaluation of methane dry reforming process on a membrane reactor to hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fabiano S.A.; Benachour, Mohand; Abreu, Cesar A.M. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. of Chemical Engineering], Email: f.aruda@yahoo.com.br

    2010-07-01

    In a fixed bed membrane reactor evaluations of methane-carbon dioxide reforming over a Ni/{gamma}- Al{sub 2}O{sub 3} catalyst were performed at 773 K, 823 K and 873 K. A to convert natural gas into syngas a fixed-bed reactor associate with a selective membrane was employed, where the operating procedures allowed to shift the chemical equilibrium of the reaction in the direction of the products of the process. Operations under hydrogen permeation, at 873 K, promoted the increase of methane conversion, circa 83%, and doubled the yield of hydrogen production, when compared with operations where no hydrogen permeation occurred. (author)

  19. Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

    Science.gov (United States)

    Damle, Ashok S.

    One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and

  20. Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

    Science.gov (United States)

    Bagni, T.; Duchateau, J. L.; Breschi, M.; Devred, A.; Nijhuis, A.

    2017-09-01

    Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability margin of magnets. In the last decade ITER has launched a campaign for characterization of several types of NbTi and Nb3Sn CICCs comprising quench tests with a singular sine wave fast magnetic field pulse and relatively small amplitude. The stability tests, performed in the SULTAN facility, were reproduced and analyzed using two codes: JackPot-AC/DC, an electromagnetic-thermal numerical model for CICCs, developed at the University of Twente (van Lanen and Nijhuis 2010 Cryogenics 50 139-148) and multi-constant-model (MCM) (Turck and Zani 2010 Cryogenics 50 443-9), an analytical model for CICCs coupling losses. The outputs of both codes were combined with thermal, hydraulic and electric analysis of superconducting cables to predict the minimum quench energy (MQE) (Bottura et al 2000 Cryogenics 40 617-26). The experimental AC loss results were used to calibrate the JackPot and MCM models and to reproduce the energy deposited in the cable during an MQE test. The agreement between experiments and models confirm a good comprehension of the various CICCs thermal and electromagnetic phenomena. The differences between the analytical MCM and numerical JackPot approaches are discussed. The results provide a good basis for further investigation of CICC stability under plasma scenario conditions using magnetic field pulses with lower ramp rate and higher amplitude.

  1. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  2. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  3. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  4. Experimental analysis of arsenic precipitation during microbial sulfate and iron reduction in model aquifer sediment reactors

    Science.gov (United States)

    Kirk, Matthew F.; Roden, Eric E.; Crossey, Laura J.; Brealey, Adrian J.; Spilde, Michael N.

    2010-05-01

    Microbial SO 42- reduction limits accumulation of aqueous As in reducing aquifers where the sulfide that is produced forms minerals that sequester As. We examined the potential for As partitioning into As- and Fe-sulfide minerals in anaerobic, semi-continuous flow bioreactors inoculated with 0.5% (g mL -1) fine-grained alluvial aquifer sediment. A fluid residence time of three weeks was maintained over a ca. 300-d incubation period by replacing one-third of the aqueous phase volume of the reactors with fresh medium every seven days. The medium had a composition comparable to natural As-contaminated groundwater with slightly basic pH (7.3) and 7.5 μM aqueous As(V) and also contained 0.8 mM acetate to stimulate microbial activity. Medium was delivered to a reactor system with and without 10 mmol L -1 synthetic goethite (α-FeOOH). In both reactors, influent As(V) was almost completely reduced to As(III). Pure As-sulfide minerals did not form in the Fe-limited reactor. Realgar (As 4S 4) and As 2S 3(am) were undersaturated throughout the experiment. Orpiment (As 2S 3) was saturated while sulfide content was low (˜50 to 150 μM), but precipitation was likely limited by slow kinetics. Reaction-path modeling suggests that, even if these minerals had formed, the dissolved As content of the reactor would have remained at hazardous levels. Mackinawite (Fe 1 + xS; x ⩽ 0.07) formed readily in the Fe-bearing reactor and held dissolved sulfide at levels below saturation for orpiment and realgar. The mackinawite sequestered little As (<0.1 wt.%), however, and aqueous As accumulated to levels above the influent concentration as microbial Fe(III) reduction consumed goethite and mobilized adsorbed As. A relatively small amount of pyrite (FeS 2) and greigite (Fe 3S 4) formed in the Fe-bearing reactor when we injected a polysulfide solution (Na 2S 4) to a final concentration of 0.5 mM after 216, 230, 279, and 286 days. The pyrite, and to a lesser extent the greigite, that formed

  5. COMPUTATIONAL AND EXPERIMENTAL MODELING OF THREE-PHASE SLURRY-BUBBLE COLUMN REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Isaac K. Gamwo; Dimitri Gidaspow

    1999-09-01

    Considerable progress has been achieved in understanding three-phase reactors from the point of view of kinetic theory. In a paper in press for publication in Chemical Engineering Science (Wu and Gidaspow, 1999) we have obtained a complete numerical solution of bubble column reactors. In view of the complexity of the simulation a better understanding of the processes using simplified analytical solutions is required. Such analytical solutions are presented in the attached paper, Large Scale Oscillations or Gravity Waves in Risers and Bubbling Beds. This paper presents analytical solutions for bubbling frequencies and standing wave flow patterns. The flow patterns in operating slurry bubble column reactors are not optimum. They involve upflow in the center and downflow at the walls. It may be possible to control flow patterns by proper redistribution of heat exchangers in slurry bubble column reactors. We also believe that the catalyst size in operating slurry bubble column reactors is not optimum. To obtain an optimum size we are following up on the observation of George Cody of Exxon who reported a maximum granular temperature (random particle kinetic energy) for a particle size of 90 microns. The attached paper, Turbulence of Particles in a CFB and Slurry Bubble Columns Using Kinetic Theory, supports George Cody's observations. However, our explanation for the existence of the maximum in granular temperature differs from that proposed by George Cody. Further computer simulations and experiments involving measurements of granular temperature are needed to obtain a sound theoretical explanation for the possible existence of an optimum catalyst size.

  6. Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

    Science.gov (United States)

    Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

    2017-12-01

    The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

  7. Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

    Science.gov (United States)

    Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

    1994-07-01

    Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding.

  8. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  9. Theoretical and Experimental Evaluation of the Temperature Distribution in a Dry Type Air Core Smoothing Reactor of HVDC Station

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2017-05-01

    Full Text Available The outdoor ultra-high voltage (UHV dry-type air-core smoothing reactors (DASR of High Voltage Direct Current systems are equipped with a rain cover and an acoustic enclosure. To study the convective heat transfer between the DASR and the surrounding air, this paper presents a coupled model of the temperature and fluid field based on the structural features and cooling manner. The resistive losses of encapsulations calculated by finite element method (FEM were used as heat sources in the thermal analysis. The steady fluid and thermal field of the 3-D reactor model were solved by the finite volume method (FVM, and the temperature distribution characteristics of the reactor were obtained. Subsequently, the axial and radial temperature distributions of encapsulation were investigated separately. Finally, an optical fiber temperature measurement scheme was used for an UHV DASR under natural convection conditions. Comparative analysis showed that the simulation results are in good agreement with the experimental data, which verifies the rationality and accuracy of the numerical calculation. These results can serve as a reference for the optimal design and maintenance of UHV DASRs.

  10. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  11. Experimental studies on catalytic hydrogen recombiners for light water reactors; Experimentelle Untersuchungen zu katalytischen Wasserstoffkombinatoren fuer Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Drinovac, P.

    2006-06-19

    In the course of core melt accidents in nuclear power plants a large amount of hydrogen can be produced and form an explosive or even detonative gas mixture with aerial oxygen in the reactor building. In the containment atmosphere of pressurized water reactors hydrogen combines a phlogistically with the oxygen present to form water vapor even at room temperature. In the past, experimental work conducted at various facilities has contributed little or nothing to an understanding of the operating principles of catalytic recombiners. Hence, the purpose of the present study was to conduct detailed investigations on a section of a recombiner essentially in order to deepen the understanding of reaction kinetics and heat transport processes. The results of the experiments presented in this dissertation form a large data base of measurements which provides an insight into the processes taking place in recombiners. The reaction-kinetic interpretation of the measured data confirms and deepens the diffusion theory - proposed in an earlier study. Thus it is now possible to validate detailed numeric models representing the processes in recombiners. Consequently the present study serves to broaden and corroborate competence in this significant area of reactor technology. In addition, the empirical knowledge thus gained may be used for a critical reassessment of previous numeric model calculations. (orig.)

  12. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  13. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    to successively improving data, the repeated instantiation of a skeleton incurs a certain overhead that could be saved by reusing existing processes, threads and communication structures. This is especially important when running parallel applications in a distributed environment. However, customising......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...... a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...

  14. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    OpenAIRE

    Mark Daniel G. de Luna; Rance, Diana Pearl M.; Luzvisminda M. Bellotindos; Lu, Ming-Chun

    2016-01-01

    This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD) wa...

  15. Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

    Science.gov (United States)

    Hall, M. M., Jr.; Flinn, J. E.

    2010-01-01

    Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.

  16. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  17. Experimental Study on Ash-Returned Reactor of CFB Atmospheric Air Gasifier

    Science.gov (United States)

    Shihong, Zhang; Luning, Tian; Xianrong, Zhou; Hanping, Chen; Haiping, Yang; Xianhua, Wang

    In an attempt to improve the gasification efficiency and decrease the carbon content in fly ash of atmospheric air CFB gasifiers, an innovatory equipment by name ash-returned reactor is put forward by SKLCC. Ash-returned reactor is an ash-returned apparatus on line of ash circulation, typically like "U" type valve in CFB boilers, with additional function of some extent combustion of residual carbon and increase the furnace inlet temperature of returning ash, and hence the coal conversion of gasifiers is enhanced. As to its configuration compared to conventional "U" type valve, ash-returned rector has two distinguished features of several times of height scale of fluidizing transportation region to meet the combustion reaction time need and appropriate heat transfer tube bundles arranged in the region to moderate the local temperature so as to avoid slagging. And hence, corresponding to the structure renovation, the material transportation and regulation performance of ash-returned reactor is primarily investigated through a series of experiments in a cold lab-scale facility in this paper. The heat transfer characteristic of the tube bundles is then researched and its influential factors are further discussed. These works lay a foundation on the following study of hot state experiments and industrial applications.

  18. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E.; Andersson, T.L.; Brunfelter, B.; Kockum, J.; Londen, S.O.; Tiren, L.I.

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

  19. Experimental evaluation of iterative reconstruction versus filtered back projection for 3D [O-15]water PET activation studies using statistical parametric mapping analysis.

    NARCIS (Netherlands)

    Mesina, C.T.; Boellaard, R.; Jongbloed, G.; van der Vaart, A.W.; Lammertsma, A.A.

    2003-01-01

    Iterative reconstructions are increasingly used for clinical PET studies owing to the better noise properties compared with filtered backprojection (FBP). The purpose of the present study was to compare ordered subsets expectation maximization (OSEM) iterative reconstruction with FBP as input for

  20. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed and Entrained-Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A. [Univ. of Utah, Salt Lake City, UT (United States); Morrill, Mike [Univ. of Utah, Salt Lake City, UT (United States); Lighty, JoAnn S. [Univ. of Utah, Salt Lake City, UT (United States); Silcox, Geoffrey D. [Univ. of Utah, Salt Lake City, UT (United States)

    2009-06-01

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  1. Experimental studies into the thermal-hydraulic performance of the VK-300 reactor based on a draft tube model

    Directory of Open Access Journals (Sweden)

    N.P. Serdun

    2015-12-01

    Full Text Available The paper presents an experimental study into the thermal-hydraulic performance of the VK-300 reactor based on a model of a single draft tube at a pressure of 3.4MPa, various flow rates and the model inlet relative enthalpies of –0.05 to 0.2. The experimental procedures include generation of a steam-water mixture circulation with a preset flow rate and a relative enthalpy through the test section at a pressure of 3.3 to 3.4MPa, and measurement of thermal-hydraulic parameters within the circuit's representative upflow and downflow lengths of practical interest. There have been confirmed the designs used to support the reactor facility serviceability and the assumptions concerning the thermal-hydraulic performance of a natural circulation circuit used in the analysis thereof. It has been shown that, across the analyzed range of the relative enthalpy values, the draft tube has an annular-dispersed or an annular flow of the steam-water mixture, both providing for the significant separation of the steam-water mixture (Ksep=0.4 at the draft tube edges and in the mixing chamber. The perforation in the upper part of the draft tubes allows the separation coefficient to be increased at the first stage and creates more favorable conditions for the second-stage separation. The measured values of the void fraction in the mixing chamber and in the draft tube are in a satisfactory agreement with calculations based on Z.L. Miropolskiy's method and the RELAP code and may be used to verify the VK-300 thermal-hydraulic codes. It has been shown that steam may enter the ring slit that simulates the annular space and reach the reactor core inlet. Further investigations need to be conducted to study this effect for its guaranteed exclusion and for the development of emergency response procedures.

  2. Development and verification test of integral reactor major components - Development of MCP impeller design, performance prediction code and experimental verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Myung Kyoon; Oh, Woo Hyoung; Song, Jae Wook [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    The present study is aimed at developing a computational code for design and performance prediction of an axial-flow pump. The proposed performance prediction method is tested against a model axial-flow pump streamline curvature method. The preliminary design is made by using the ideal velocity triangles at inlet and exit and the three dimensional blade shape is calculated by employing the free vortex design method. Then the detailed blading design is carried out by using experimental database of double circular arc cambered hydrofoils. To computationally determine the design incidence, deviation, blade camber, solidity and stagger angle, a number of correlation equations are developed form the experimental database and a theorical formula for the lift coefficient is adopted. A total of 8 equations are solved iteratively using an under-relaxation factor. An experimental measurement is conducted under a non-cavitating condition to obtain the off-design performance curve and also a cavitation test is carried out by reducing the suction pressure. The experimental results are very satisfactorily compared with the predictions by the streamline curvature method. 28 refs., 26 figs., 11 tabs. (Author)

  3. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  4. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  5. Experimental demonstration of iterative post-equalization algorithm for 37.5-Gbaud PM-16QAM quad-carrier Terabit superchannel.

    Science.gov (United States)

    Jia, Zhensheng; Chien, Hung-Chang; Cai, Yi; Yu, Jianjun; Zhang, Chengliang; Li, Junjie; Ma, Yiran; Shang, Dongdong; Zhang, Qi; Shi, Sheping; Wang, Huitao

    2015-02-09

    We experimentally demonstrate a quad-carrier 1-Tb/s solution with 37.5-GBaud PM-16QAM signal over 37.5-GHz optical grid at 6.7 b/s/Hz net spectral efficiency. Digital Nyquist pulse shaping at the transmitter and post-equalization at the receiver are employed to mitigate the impairments of joint inter-symbol-interference (ISI) and inter-channel-interference (ICI) symbol degradation. The post-equalization algorithms consist of one sample/symbol based decision-directed least mean square (DD-LMS) adaptive filter, digital post filter and maximum likelihood sequence estimation (MLSE), and a positive iterative process among them. By combining these algorithms, the improvement as much as 4-dB OSNR (0.1nm) at SD-FEC limit (Q(2) = 6.25 corresponding to BER = 2.0e-2) is obtained when compared to no such post-equalization process, and transmission over 820-km EDFA-only standard single-mode fiber (SSMF) link is achieved for two 1.2-Tb/s signals with the averaged Q(2) factor larger than 6.5 dB for all sub-channels. Additionally, 50-GBaud 16QAM operating at 1.28 samples/symbol in a DAC is also investigated and successful transmission over 410-km SSMF link is achieved at 62.5-GHz optical grid.

  6. Experimental investigation and model validation of a CaO/Ca(OH)2 fluidized bed reactor for thermochemical energy storage applications

    OpenAIRE

    Álvarez Criado, Yolanda; Huille, Alfred; Rougé, Sylvie; Abanades García, Juan Carlos

    2016-01-01

    The CaO/Ca(OH)2 hydration/dehydration chemical loop has long been recognized as a potential candidate for application in energy storage systems for concentrated solar plants. However, the technology still remains at a conceptual level because little information has been published on the performance of the key reactors in the system. In this work, we experimentally investigate the hydration and dehydration reactors in a 5.5 kWth batch fluidized bed reactor, in conditions relevant to larger sys...

  7. Water hydraulic actuators for ITER maintenance devices

    Energy Technology Data Exchange (ETDEWEB)

    Siuko, Mikko E-mail: mikko.siuko@tut.fi; Pitkaeaho, M.; Raneda, A.; Poutanen, J.; Tammisto, J.; Palmer, J.; Vilenius, M

    2003-09-01

    The characteristic advantages of hydraulics (high power density, simple construction and reliability) together with the characteristics of water as the pressure medium (fire and environmentally safe, chemically neutral, not activated nor affected by radiation) are highlighted in critical applications such as remote handling operations in international thermonuclear experimental reactor (ITER). However, lack of commercial selection of water hydraulic components, common design expertise and known application experiences prevents wide use of water hydraulics. Since 1994, IHA has designed and manufactured water hydraulic tools for ITER divertor maintenance and experiences have been good. Therefore, IHA is developing water hydraulic component selection to be applied in coming systems where water hydraulics is foreseen to provide an advantage. Aim of the still on going project is to develop a set of components like power units, control components and actuators. By that way designers are able to apply water hydraulics where advantageous. In the paper the component types, their design and characteristics and results obtained so far are presented.

  8. Feasibility Study of a Novel Membrane Reactor for Syngas Production. Part 1: Experimental Study of O2 Permeation through Perovskite Membranes under Reducing and Non-Reducing Atmospheres

    NARCIS (Netherlands)

    Zhang Wenxing, Z.W.; Zhang, Wenxing; Smit, J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    In this contribution, the feasibility of a novel membrane reactor for energy efficient syngas production is investigated by means of an experimental and a simulation study. In Part 1, a detailed experimental study is performed on the O2 permeation through a perovskite membrane with composition

  9. Selective hydrogenation in trickle-bed reactor. Experimental and modelling including partial wetting.

    OpenAIRE

    Dietz, Adrian; Julcour-Lebigue, Carine; Wilhelm, Anne-Marie; Delmas, Henri

    2003-01-01

    International audience; A steady state model of a trickle bed reactor is developed for the consecutive hydrogenation of 1,5,9-cyclododecatriene on a Pd/Al2O3 catalyst. Various experiments have shown that the selectivity of this reaction towards the product of interest is much lower in co-current down-flow (trickle-bed) than in up-flow. This is due to uneven liquid distribution and to partial wetting of the catalyst surface at low liquid flow rates. The non-isothermal heterogeneous model propo...

  10. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias

    2005-07-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  11. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  12. Experimental study of hydrodynamic and operation start of a baffled anaerobic reactor treating sewage

    Directory of Open Access Journals (Sweden)

    Ana Carolina Silveira Perico

    2009-12-01

    Full Text Available It is important to provide individual sanitation systems for sewage peri-urban communities or rural areas to minimize impacts on the environment and human health caused by sewage discharge in natura into water resources. In this context, the anaerobic digestion of effluent has been one of the main considered technologies due to easy implementation, material minimization and reduction in waste production. The objective of this work was to study a Baffled Anaerobic Reactor (BAR including its hydrodynamic characteristics, percentile of inoculum to be applied and reactor operation start. It was concluded that the flow is dispersed with 3.84% of dead spaces and that 20% of the cow manure provided best results; however, due to the high fiber content of the manure, its use is not recommended as inoculum. The BAR system, composed of four chambers, presented good performance for sewage treatment of a rural community in terms of organic substance removal (COD, turbidity and solids meeting effluent disposal standards of these parameters considering the Federal and Minas Gerais State legislation, in Brazil, even in a transient phase of operation, at temperatures below 20°C. However, the effluents from the BAR can’t be released into water bodies without other parameters such as nitrogen, phosphorus, fecal coliforms, and others are investigated to be conforming to those standards.

  13. Theoretical and experimental study of the photocatalytic activity of ZnO coated tubular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ríos-Valdovinos, E.; Amézaga-Madrid, P.; Antúnez-Flores, W.; Pola-Albores, F.; Pizá-Ruiz, P.; Miki-Yoshida, M., E-mail: mario.miki@cimav.edu.mx

    2014-01-15

    Highlights: • High quality ZnO thin films were deposited on the internal surface of fused silica tubing. • Surface carrier concentration was calculated theoretically under external irradiation. • Influence of film thickness on photocatalytic activity was explained by this model. • An optimum thickness around 60–70 nm was determined to get highest activity. -- Abstract: ZnO thin films were deposited inside of fused silica tubing by aerosol assisted chemical vapor deposition technique. The films were transparent, uniform, highly adherent and non-light scattering. Photocatalytic activity of internally ZnO coated tubing was evaluated by discoloration of a methyl orange aqueous solution in a batch reactor. Tubing was externally irradiated with UV-A at room temperature. A one dimensional model was proposed to calculate the spatial distribution of the carrier density and the films’ surface charge carrier concentration. This model can explain the influence of the films thickness on the photocatalytic activity. Results showed that the photocatalytic activity largely depends on the film thickness. For external irradiation of the films the optimum thickness was around 60–70 nm, for which the photocatalytic activity was maximum. The photonic efficiency of internally ZnO coated tubular reactors was evaluated as a function of initial colorant concentration, irradiation time and intensity. Furthermore, due to the high activity of the ZnO films, the films were repeatedly exposed to UV-A irradiation cycles, followed by activity measurement.

  14. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  15. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  16. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission, Joint Research Center, Karlsruhe (Germany). Inst. for Transuranium Elements; Avincola, V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. fuer Angewandte Materialien (IAM-AWP); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-11-15

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRi-ISOtropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1,600 to 1,650 C, remaining well below the melting point of the fuel. Two key aspects associated with the safety of HTR fuel are assessed in this paper: fission product retention at temperatures up to 1,800 C is analyzed with the Cold Finger Apparatus (KueFA) while the behaviour of HTR-relevant fuel materials in an oxidizing environment is studied with the Corrosion Apparatus KORA. The KueFA is used to observe the combined effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in helium atmosphere for several hundred hours, mimicking accident temperatures up to 1,800 C and realistic temperature transients. Nongaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium (HPGe) gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. To achieve a good quantification of the release, a careful calibration of the setup is necessary and a collimator needs to be used in some cases. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Another relevant HTR accident scenario is air ingress into the reactor vessel as a consequence of a DLOFC incident. In

  17. Experimental investigation of the solar carbothermic reduction of ZnO using a two-cavity solar reactor

    Energy Technology Data Exchange (ETDEWEB)

    Frommherz, U.; Osinga, T.; Steinfeld, A.; Wieckert, C.

    2003-03-01

    Zinc production by solar carbothermic reduction of ZnO offers a CO{sub 2} emission reduction by a factor of 5 vis-a-vis the conventional fossil-fuel-based electrolytic or Imperial Smelting processes. Zinc can serve as a fuel in Zn-air fuel cells or can be further reacted with H{sub 2}O to form high-purity H{sub 2}. In either case, the product ZnO can be solar-recycled to Zn. We report on experimental results obtained with a 5 kW solar chemical reactor prototype that features two cavities in series, with the inner one functioning as the solar absorber and the outer one as the reaction chamber. Tests were conducted at PSI's Solar Furnace and ETH's High-Flux Solar Simulator to investigate the effect of process temperature (range 1350-1600 K) and reducing agent type (beech charcoal, activated charcoal, petcoke) on the reactor's performance and on the chemical conversion. In a typical 40-min solar experiment at 1500 K, 500 g of a 1:0.8 stoichiometric ZnO-C mixture were processed into Zn(g), CO, and CO{sub 2}. Thermal efficiencies of up to 20% were achieved. (author)

  18. Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pitman, Stan G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Good, Morris S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walker, Cody M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-07-16

    The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

  19. Project of law, adopted by the Senate, giving permission to the approval of the agreement between the French government and the international organization for thermonuclear fusion energy ITER, relative to the head office of ITER organization and to the privileges and immunities of ITER organization in the French territory; Projet de loi adopte par le Senat, autorisant l'approbation de l'accord entre le Gouvernement de la Republique francaise et l'Organisation internationale ITER pour l'energie de fusion relatif au siege de l'Organisation ITER et aux privileges et immunites de l'Organisation ITER sur le territoire francais

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-01-15

    The will of building up an international thermonuclear experimental reactor (ITER) gathers since several years the European community of atomic energy (Euratom), Japan, the USA, and Russia, next followed by China, South Korea and, since 2005, by India. The agreement signed in Paris between these seven parties on November 21, 2006 entrusted the international organization ITER with the realization of this project. The implications of the ITER project are enormous both in their scientific and in their economical aspects. France has a particular position in this project since the head office of ITER organisation is sited at Saint-Paul-lez-Durance and the tokamak will be built at Cadarache. Therefore, an agreement has been signed between ITER organization and the French government. The approval of this agreement is the object of this project of law made of a single article. The agreement between the French government and the international organization ITER is attached to the document. It defines the juridical status, the privileges and immunities of the organization itself and of its personnel inside the French territory. An appendix to the agreement precises the cooperation modalities between the French authorities and ITER organization. (J.S.)

  20. Combined numerical and experimental investigation into the coolant flow hydrodynamics and mass transfer behind the spacer grid in fuel assemblies of the floating power unit reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-09-01

    Full Text Available The results of experimental investigations into the local hydrodynamics and inter-cell mass transfer of the coolant flow in representative zones of the KLT-40C reactor FAs behind the plate-type spacer grid are presented. The investigations were conducted on an aerodynamic rig using the admixture diffusion method (the tracer-gas method. A study into the spatial dispersion of the absolute flow velocity projections and into the distribution of the tracer concentration allowed specify the coolant flow pattern behind the FA plate-type spacer grid of the KLT-40C reactor. The results of measuring the flow friction coefficient in the plate-type spacer grid, depending on the Reynolds number, are presented. Based on the obtained experimental data, recommendations have been provided for updating the procedures to calculate the coolant flow rates for the KLT-40C reactor core by-channel codes. The results of investigating the coolant flow local hydrodynamics and mass transfer in the KLT-40C reactor FAs have been adopted for practical use by Afrikantov OKBM for estimating the heat-engineering reliability of the KLT-40C reactor cores and have been data based for verification of CFD codes and detailed by-channel calculation of the KLT-40C reactor core.

  1. Iterative participatory design

    DEFF Research Database (Denmark)

    Simonsen, Jesper; Hertzum, Morten

    2010-01-01

    iterative process of mutual learning by designers and domain experts (users), who aim to change the users’ work practices through the introduction of information systems. We provide an illustrative case example with an ethnographic study of clinicians experimenting with a new electronic patient record......The theoretical background in this chapter is information systems development in an organizational context. This includes theories from participatory design, human-computer interaction, and ethnographically inspired studies of work practices. The concept of design is defined as an experimental...

  2. Monitoring of dry anaerobic fermentation in experimental facility with use of biofilm reactor

    Directory of Open Access Journals (Sweden)

    Milan Šinkora

    2011-01-01

    -called biofilm reactor. An external reactor with a cultivated bacterial biofilm on an immovable carrier with the percolate flowing through it has been constructed in laboratory conditions for this purpose. The choice of suitable percolate strategy (this means the frequency of sprinkling and the amount of percolate directly influences the process of anaerobic fermentation.

  3. Experimental Investigations of Physical and Chemical Properties for Microalgae HTL Bio-Crude Using a Large Batch Reactor

    Directory of Open Access Journals (Sweden)

    Farhad M. Hossain

    2017-04-01

    Full Text Available As a biofuel feedstock, microalgae has good scalability and potential to supply a significant proportion of world energy compared to most types of biofuel feedstock. Hydrothermal liquefaction (HTL is well-suited to wet biomass (such as microalgae as it greatly reduces the energy requirements associated with dewatering and drying. This article presents experimental analyses of chemical and physical properties of bio-crude oil produced via HTL using a high growth-rate microalga Scenedesmus sp. in a large batch reactor. The overarching goal was to investigate the suitability of microalgae HTL bio-crude produced in a large batch reactor for direct application in marine diesel engines. To this end we characterized the chemical and physical properties of the bio-crudes produced. HTL literature mostly reports work using very small batch reactors which are preferred by researchers, so there are few experimental and parametric measurements for bio-crude physical properties, such as viscosity and density. In the course of this study, a difference between traditionally calculated values and measured values was noted. In the parametric study, the bio-crude viscosity was significantly closer to regular diesel and biodiesel standards than transesterified (FAME microalgae biodiesel. Under optimised conditions, HTL bio-crude’s high density (0.97–1.04 kg·L−1 and its high viscosity (70.77–73.89 mm2·s−1 had enough similarity to marine heavy fuels. although the measured higher heating value, HHV, was lower (29.8 MJ·kg−1. The reaction temperature was explored in the range 280–350 °C and bio-crude oil yield and HHV reached their maxima at the highest temperature. Slurry concentration was explored between 15% and 30% at this temperature and the best HHV, O:C, and N:C were found to occur at 25%. Two solvents (dichloromethane and n-hexane were used to recover the bio-crude oil, affecting the yield and chemical composition of the bio-crude.

  4. Experimental investigation of pebble flow dynamics using radioactive particle tracking technique in a scaled-down Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Khane, Vaibhav; Said, I.A.; Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu

    2016-06-15

    Highlights: • Pebble Flow fields at Pebble Bed Modular Reactor was investigated. • Radioactive Particle Tracking (RPT) technique has been used. • Plug flow type velocity profile is suggested at upper cylindrical region. - Abstract: The Pebble Bed Modular Reactor (PBMR) is a type of very-high-temperature reactor (VHTR) that is conceptually very similar to moving bed reactors used in the chemical and petrochemical industries. In a PBMR core, nuclear fuel is in the form of pebbles and moves slowly under the influence of gravity. In this work, an integrated experimental and computational study of granular flow in a scaled-down cold flow PBMR was performed. A continuous pebble re-circulation experimental set-up, mimicking the flow of pebbles in a PBMR was designed and developed. An experimental investigation of pebble flow dynamics in a scaled down test reactor was carried out using a non-invasive radioactive particle tracking (RPT) technique that used a cobalt-60 based tracer to mimic pebbles in terms of shape, size and density. A cross-correlation based position reconstruction algorithm and RPT calibration data were used to obtain results about Lagrangian trajectories, the velocity field, and residence time distributions. The RPT technique results a serve as a benchmark data for assessing contact force models used in the discrete element method (DEM) simulations.

  5. Flow balancing orifice for ITER toroidal field coil

    Science.gov (United States)

    Litvinovich, A. V.; Y Rodin, I.; Kovalchuk, O. A.; Safonov, A. V.; Stepanov, D. B.; Guryeva, T. M.

    2017-12-01

    Flow balancing orifices (FBOs) are used in in International thermonuclear experimental reactor (ITER) Toroidal Field coil to uniform flow rate of cooling gas in the side double pancakes which have a different conductor length: 99 m and 305 m, respectively. FBOs consist of straight parts, elbows produced from a 316L stainless steel tube 21.34 x 2.11 mm and orifices made from a 316L stainless steel rod. Each of right and left FBOs contains 6 orifices, straight FBOs contain 4 and 6 orifices. Before manufacturing of qualification samples D.V. Efremov Institute of Electrophysical Apparatus (JSC NIIEFA) proposed to ITER a new approach to provide the seamless connection between a tube and a plate therefore the most critical weld between the orifice with 1 mm thickness and the tube removed from the FBOs final design. The proposed orifice diameter is three times less than the minimum requirement of the ISO 5167, therefore it was tasked to define accuracy of calculation flow characteristics at room temperature and compare with the experimental data. In 2015 the qualification samples of flow balancing orifices were produced and tested. The results of experimental data showed that the deviation of calculated data is less than 7%. Based on this result and other tests ITER approved the design of FBOs, which made it possible to start the serial production. In 2016 JSC NIIEFA delivered 50 FBOs to ITER, i.e. 24 left side, 24 right side and 2 straight FBOs. In order to define the quality of FBOs the test facility in JSC NIIEFA was prepared. The helium tightness test at 10-9 m3·Pa/s the pressure up to 3 MPa, flow rate measuring at the various pressure drops, the non-destructive tests of orifices and weld seams (ISO 5817, class B) were conducted. Other tests such as check dimensions and thermo cycling 300 - 80 - 300 K also were carried out for each FBO.

  6. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  7. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Science.gov (United States)

    Buckingham, Grant T.; Ormond, Thomas K.; Porterfield, Jessica P.; Hemberger, Patrick; Kostko, Oleg; Ahmed, Musahid; Robichaud, David J.; Nimlos, Mark R.; Daily, John W.; Ellison, G. Barney

    2015-01-01

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C6H5CH2, as well as a set of isotopically labeled radicals: C6H5CD2, C6D5CH2, and C6H513CH2. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C5H4=C=CH2, H atom, C5H4—C ≡ CH, C5H5, HCCCH2, and HC ≡ CH. Pyrolysis of the C6H5CD2, C6D5CH2, and C6H513CH2 benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C6H5CH2⇋C7H7. These labeling studies suggest that there must be other thermal decomposition routes for the C6H5CH2 radical that differ from the fulvenallene pathway.

  8. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  9. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    Energy Technology Data Exchange (ETDEWEB)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  10. Change of interstrand contact resistance and coupling loss in various prototype ITER NbTi conductors with transverse loading in the Twente Cryogenic Cable Press up to 40,000 cycles

    NARCIS (Netherlands)

    Nijhuis, Arend; Ilyin, Y.; Abbas, Wouter; ten Haken, Bernard; ten Kate, Herman H.J.

    2004-01-01

    The multistrand NbTi conductors for the Poloidal Field (PF) Coils of the International Thermonuclear Experimental Reactor (ITER) are subjected to heavy transverse loading due to the Lorentz forces in the coils. The current in the multistage Cable-In-Conduit Conductors (CICC) exceeds 50 kA and the

  11. The role of clusters in gas-solids reactors. An experimental study.

    NARCIS (Netherlands)

    Venderbosch, R.H.

    1998-01-01

    This PhD-work is meant to determine the contact efficiency experimentally for fluidization of fine particles over a wide range of superficial gas velocities (dp<200 mm and 0.1

  12. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Grant T.; Ormond, Thomas K. [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Porterfield, Jessica P.; Ellison, G. Barney [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); Hemberger, Patrick [Molecular Dynamics Group, Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Kostko, Oleg; Ahmed, Musahid [Chemical Sciences Division, Lawrence Berkeley National Laboratories, Berkeley, California 94720 (United States); Robichaud, David J.; Nimlos, Mark R. [National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Daily, John W. [Department of Mechanical Engineering, Center for Combustion and Environmental Research,University of Colorado, Boulder, Colorado 80309-0427 (United States)

    2015-01-28

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C{sub 6}H{sub 5}CH{sub 2}, as well as a set of isotopically labeled radicals: C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C{sub 5}H{sub 4}=C=CH{sub 2}, H atom, C{sub 5}H{sub 4}—C ≡ CH, C{sub 5}H{sub 5}, HCCCH{sub 2}, and HC ≡ CH. Pyrolysis of the C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2} benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C{sub 6}H{sub 5}CH{sub 2}⇋C{sub 7}H{sub 7}. These labeling studies suggest that there must be other thermal decomposition routes for the C{sub 6}H{sub 5}CH{sub 2} radical that differ from the fulvenallene pathway.

  13. Three-phase packed bed reactor with an evaporating solvent—I. Experimental: the hydrogenation of 2,4,6-trinitrotoluene in methanol

    NARCIS (Netherlands)

    van Gelder, K.B.; Damhof, J.K.; Kroijenga, P.J.; Westerterp, K.R.

    1990-01-01

    In this paper we present experimental data on the three-phase hydrogenation of 2,4,6-trinitrotoluene (TNT) to triaminotoluene. The experiments are performed in a cocurrent upflow packed bed reactor. Methanol is used as an evaporating solvent. The influence of the main operating parameters, the

  14. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  15. An Experimental Investigation of Sewage Sludge Gasification in a Fluidized Bed Reactor

    Science.gov (United States)

    Calvo, L. F.; García, A. I.; Otero, M.

    2013-01-01

    The gasification of sewage sludge was carried out in a simple atmospheric fluidized bed gasifier. Flow and fuel feed rate were adjusted for experimentally obtaining an air mass : fuel mass ratio (A/F) of 0.2 gasification. This allowed improving the process heat transfer and, therefore, gasification efficiency. The heating value of the produced gas was 8.4 MJ/Nm, attaining a hot gas efficiency of 70% and a cold gas efficiency of 57%. PMID:24453863

  16. Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

    NARCIS (Netherlands)

    Bagni, T.; Duchateau, J.L.; Breschi, M.; Devred, A.; Nijhuis, A.

    2017-01-01

    Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability

  17. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  18. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  19. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  20. Environmental dose rate assessment of ITER using the Monte Carlo method

    Directory of Open Access Journals (Sweden)

    Karimian Alireza

    2014-01-01

    Full Text Available Exposure to radiation is one of the main sources of risk to staff employed in reactor facilities. The staff of a tokamak is exposed to a wide range of neutrons and photons around the tokamak hall. The International Thermonuclear Experimental Reactor (ITER is a nuclear fusion engineering project and the most advanced experimental tokamak in the world. From the radiobiological point of view, ITER dose rates assessment is particularly important. The aim of this study is the assessment of the amount of radiation in ITER during its normal operation in a radial direction from the plasma chamber to the tokamak hall. To achieve this goal, the ITER system and its components were simulated by the Monte Carlo method using the MCNPX 2.6.0 code. Furthermore, the equivalent dose rates of some radiosensitive organs of the human body were calculated by using the medical internal radiation dose phantom. Our study is based on the deuterium-tritium plasma burning by 14.1 MeV neutron production and also photon radiation due to neutron activation. As our results show, the total equivalent dose rate on the outside of the bioshield wall of the tokamak hall is about 1 mSv per year, which is less than the annual occupational dose rate limit during the normal operation of ITER. Also, equivalent dose rates of radiosensitive organs have shown that the maximum dose rate belongs to the kidney. The data may help calculate how long the staff can stay in such an environment, before the equivalent dose rates reach the whole-body dose limits.

  1. Contribution to modeling of the reflooding of a severely damaged reactor core using PRELUDE experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, A.; Fichot, F.; Repetto, G. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France); Quintard, M. [Universite de Toulouse, INPT, UPS, IMFT Institut de Mecanique des Fluides de Toulouse, Allee Camille Soula, F-31400 Toulouse (France); CNRS, IMFT, F-31400 Toulouse (France); Fleurot, J. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France)

    2012-07-01

    In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)

  2. Transient thermal hydraulic modeling and analysis of ITER divertor plate system

    Energy Technology Data Exchange (ETDEWEB)

    El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

    2009-12-15

    A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

  3. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  4. Simulations of Alpha Wall Load in ITER. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan

    2010-10-20

    The partially DOE funded International Thermonuclear Experimental Reactor (ITER) will produce massive amounts of energetic charged alpha particles, which are imperfectly confined by a strong magnetic field. The wall of the experiment is designed to withstand an estimated wall load from these fusion alpha particles, but the accuracy of this estimate needs to be improved to avoid potentially catastrophic surprises when the experiment becomes operational. We have added a more accurate, gyro-dynamic model of particle motion to the existing drift-dynamic model in the DELTA5D simulation software used for the project. We have also added the ability to load a detailed engineering model of the wall and use it in the simulations.

  5. Conceptual design of cesium removal device for ITER NBI maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Shibanuma, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Cesium is required in order to generate a stable negative ion of hydrogen in an ion source of the neutral beam injector (NBI), which is one of the plasma-heating devices for International Thermonuclear Experimental Reactor (ITER). After long time operation of the NBI, the cesium deposits to the insulators supporting the electrode. Due to the deterioration of the insulation resistance, the continuous operation of the NBI will be difficult. In addition, the NBI device is activated by neutrons from D-T plasma, so that periodic removal and cleaning of the cesium on the insulators by remove handling is required. A study of the cesium removal scenario and the device is therefore required considering remote handling. In this report, a cesium removal procedure and conceptual design of the cesium removal device using laser ablation technique are studied, and the feasibility of the laser ablation method is shown. (author)

  6. Exploratory study of burn time, duty factor, and fluence on ITER activation hazards

    Energy Technology Data Exchange (ETDEWEB)

    Piet, S.J.

    1992-08-01

    The safety analyses for the Conceptual Design Activity (CDA) of the International Thermonuclear Experimental Reactor (ITER) were based on the simplifying assumption that the activation of materials occurs continuously. Since the analyses showed a significant hazard, it is appropriate to examine how much hazard reduction might occur if this conservative assumption were relaxed. This report explores how much reduction might be gained by considering non-continuous operation, that is, by considering plasma burn time, duty factor, and integrated fluence. Other factors impacting activation hazards - material choice, flux, and size - are not considered here.

  7. Nitritation performance and biofilm development of co- and counter-diffusion biofilm reactors: Modeling and experimental comparison

    DEFF Research Database (Denmark)

    Wang, Rongchang; Terada, Akihiko; Lackner, Susanne

    2009-01-01

    A comparative study was conducted on the start-up performance and biofilm development in two different biofilm reactors with aim of obtaining partial nitritation. The reactors were both operated under oxygen limited conditions, but differed in geometry. While substrates (O-2, NH3) co...... results showed that the counter-diffusion biofilms developed faster and attained a larger maximum biofilm thickness than the co-diffusion biofilms. Under oxygen limited condition (DO

  8. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  9. Experimental evaluation of two different types of reactors for CO2 removal from gaseous stream by bottom ash accelerated carbonation.

    Science.gov (United States)

    Lombardi, L; Carnevale, E A; Pecorini, I

    2016-12-01

    Low methane content landfill gas may be enriched by removing carbon dioxide. An innovative process, based on carbon dioxide capture and storage by means of accelerated carbonation of bottom ash is proposed and studied for the above purpose. Within this research framework we devoted a preliminary research activity to investigate the possibility of improving the way the contact between bottom ash and landfill gas takes place: this is the scope of the work reported in this paper. Two different types of reactors - fixed bed and rotating drum - were designed and constructed for this purpose. The process was investigated at laboratory scale. As the aim of this phase was the comparison of the performances of the two different reactors, we used a pure stream of CO2 to preliminarily evaluate the reactor behaviors in the most favorable condition for the process (i.e. maximum CO2 partial pressure at ambient condition). With respect to the simple fixed bed reactor concept, some modifications were proposed, consisting of separating the ash bed in three layers. With the three layer configuration we would like to reduce the possibility for the gas to follow preferential paths through the ash bed. However, the results showed that the process performances are not significantly influenced by the multiple layer arrangement. As an alternative to the fixed bed reactor, the rotating drum concept was selected in order to provide continuous mixing of the solids. Two operating parameters were considered and varied during the tests: the filling ratio and the rotating speed. Better performances were observed for lower filling ratio while the rotating speed showed minor importance. Finally the performances of the two reactors were compared. The rotating drum reactor is able to provide improved carbon dioxide removal with respect to the fixed bed one, especially when the rotating reactor is operated at low filling ratio values and slow rotating speed values. Comparing the carbon dioxide

  10. Verification test results of a cutting technique for the ITER blanket cooling pipes

    Energy Technology Data Exchange (ETDEWEB)

    Shigematsu, Soichiro, E-mail: shigematsu.soichiro@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki (Japan); Tanigawa, Hisashi; Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki (Japan); Mori, Seiji; Nakahira, Masataka; Raffray, Rene; Merola, Mario [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

    2012-08-15

    For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.

  11. Ex-vessel break in ITER divertor cooling loop analysis with the ECART code

    CERN Document Server

    Cambi, G; Parozzi, F; Porfiri, MT

    2003-01-01

    A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed.

  12. Assessment of database for interaction of tritium with ITER plasma facing materials

    Energy Technology Data Exchange (ETDEWEB)

    Dolan, T.J.; Anderl, R.A.

    1994-09-01

    The present work surveys recent literature on hydrogen isotope interactions with Be, SS and Inconels, Cu, C, and V, and alloys of Cu and V. The goals are (1) to provide input to the International Thermonuclear Experimental Reactor (ITER) team to help with tritium source term estimates for the Early Safety and Environmental Characterization Study and (2) to provide guidance for planning additional research that will be needed to fill gaps in the present materials database. Properties of diffusivity, solubility, permeability, chemical reactions, Soret effect, recombination coefficient, surface effects, trapping, porosity, layered structures, interfaces, and oxides are considered. Various materials data are tabulated, and a matrix display shows an assessment of the quality of the data available for each main property of each material. Recommendations are made for interim values of diffusivity and solubility to be used, pending further discussion by the ITER community.

  13. Vacuum Bellows, Vacuum Piping, Cryogenic Break, and Copper Joint Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2010-06-01

    The ITER international project design teams are working to produce an engineering design in preparation for construction of the International Thermonuclear Experimental Reactor (ITER) tokamak. During the course of this work, questions have arisen in regard to safety barriers and equipment reliability as important facets of system design. The vacuum system designers have asked several questions about the reliability of vacuum bellows and vacuum piping. The vessel design team has asked about the reliability of electrical breaks and copper-copper joints used in cryogenic piping. Research into operating experiences of similar equipment has been performed to determine representative failure rates for these components. The following chapters give the research results and the findings for vacuum system bellows, power plant stainless steel piping (amended to represent vacuum system piping), cryogenic system electrical insulating breaks, and copper joints.

  14. Testing of ITER prototype cable-in-conduit conductors in the FENIX facility

    Science.gov (United States)

    Shen, S. S.; Chaplin, M. R.; Felker, B.; Hassenzahl, W. V.; Kishiyama, K. I.; Parker, J. M.

    1994-07-01

    The Fusion Engineering International Experiment (FENIX) Test Facility has been operational since 1991 at the Lawrence Livermore National Laboratory for testing the International Thermonuclear Experimental Reactor (ITER) prototype conductors. These conductors are designed to operate stably with transport current of more than 40 kA at a magnetic field of 13 T. The FENIX facility consists of four magnet sets that are configured to allow easy access to the 40-cm high-field region with a test cross-section area of 10-15 cm(exp 2). FENIX provides test conditions that closely simulate many of the ITER magnet operation modes. Performed experiments include measurements of critical current, current-sharing temperature, forced-flow properties, stability, joint performance and cyclic fatigue effects. This paper describes the design and performance of these experiments.

  15. Applied iterative methods

    CERN Document Server

    Hageman, Louis A

    1981-01-01

    This graduate-level text examines the practical use of iterative methods in solving large, sparse systems of linear algebraic equations and in resolving multidimensional boundary-value problems. Assuming minimal mathematical background, it profiles the relative merits of several general iterative procedures. Topics include polynomial acceleration of basic iterative methods, Chebyshev and conjugate gradient acceleration procedures applicable to partitioning the linear system into a "red/black" block form, adaptive computational algorithms for the successive overrelaxation (SOR) method, and comp

  16. Experimental study of local coolant hydrodynamics in TVS-Kvadrat PWR reactor fuel assembly using mixing spacer grids with different types of deflectors

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-12-01

    Full Text Available Results of experimental studies of local hydrodynamic characteristics of coolant flow in fuel assemblies of RWR reactors using different types of mixing spacer grids are presented. Specific features and regularities of coolant flow in fuel pin bundles of TVS-KVADRAT fuel assemblies with different types of mixing spacer grids were revealed in the course of experiments. Analysis of space distribution of projections of absolute flow velocity allowed detailed description of coolant flow beyond the spacer grid with installation of three different types of deflectors. Optimal design of deflector for spacer grid of the TVS-KVADRAT fuel assembly in the standard cell in the area of guiding channels was identified. Results of studies of local hydrodynamics of coolant flow in the TVS-KVADRAT fuel assembly are accepted for subsequent practical application by the JSC Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM in the evaluations of thermal engineering reliability of PWR reactor cores and were included in the database for verification of computational fluid dynamic codes (CFD-codes and implementation of detailed cell array calculations of PWR reactor cores.

  17. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  18. Predictor-Corrector Quasi-Static Method Applied to Nonoverlapping Local/Global Iterations with 2-D/1-D Fusion Transport Kernel and p-CMFD Wrapper for Transient Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Bumhee; Cho, Nam Zin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, the steady-state p-CMFD adjoint flux is used as the weighting function to obtain PK parameters instead of the computationally expensive transport adjoint angular flux. Several numerical problems are investigated to see the capability of the PCQS method applied to the NLG iteration. CRX-2K adopts the nonoverlapping local/global (NLG) iterative method with the 2-D/1-D fusion transport kernel and the global p-CMFD wrapper. The parallelization of the NLG iteration has been recently implemented in CRX-2K and several numerical results are reported in a companion paper. However, the direct time discretization leads to a fine time step size to acquire an accurate transient solution, and the step size involved in the transport transient calculations is millisecond-order. Therefore, the transient calculations need much longer computing time than the steady-state calculation. To increase the time step size, Predictor-Corrector Quasi-Static (PCQS) method can be one option to apply to the NLG iteration. The PCQS method is a linear algorithm, so the shape function does not need to be updated more than once at a specific time step like a conventional quasi-static (QS) family such as Improved Quasi-Static (IQS) method. Moreover, the shape function in the PCQS method directly comes from the direct transport calculation (with a large time step), so one can easily implement the PCQS method in an existing transient transport code. Any QS method needs to solve the amplitude function in the form of the point kinetics (PK) equations, and accurate PK parameters can be obtained by the transport steady-state adjoint angular flux as a weighting function. The PCQS method is applied to the transient NLG iteration with the 2-D/1-D fusion transport kernel and the global p-CMFD wrapper, and has been implemented in CRX-2K. In the numerical problems, the PCQS method with the NLG iteration shows more accurate solutions compared to the direct transient calculations with large time step

  19. A fresh look at electron cyclotron current drive power requirements for stabilization of tearing modes in ITER

    Energy Technology Data Exchange (ETDEWEB)

    La Haye, R. J., E-mail: lahaye@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

    2015-12-10

    ITER is an international project to design and build an experimental fusion reactor based on the “tokamak” concept. ITER relies upon localized electron cyclotron current drive (ECCD) at the rational safety factor q=2 to suppress or stabilize the expected poloidal mode m=2, toroidal mode n=1 neoclassical tearing mode (NTM) islands. Such islands if unmitigated degrade energy confinement, lock to the resistive wall (stop rotating), cause loss of “H-mode” and induce disruption. The International Tokamak Physics Activity (ITPA) on MHD, Disruptions and Magnetic Control joint experiment group MDC-8 on Current Drive Prevention/Stabilization of Neoclassical Tearing Modes started in 2005, after which assessments were made for the requirements for ECCD needed in ITER, particularly that of rf power and alignment on q=2 [1]. Narrow well-aligned rf current parallel to and of order of one percent of the total plasma current is needed to replace the “missing” current in the island O-points and heal or preempt (avoid destabilization by applying ECCD on q=2 in absence of the mode) the island [2-4]. This paper updates the advances in ECCD stabilization on NTMs learned in DIII-D experiments and modeling during the last 5 to 10 years as applies to stabilization by localized ECCD of tearing modes in ITER. This includes the ECCD (inside the q=1 radius) stabilization of the NTM “seeding” instability known as sawteeth (m/n=1/1) [5]. Recent measurements in DIII-D show that the ITER-similar current profile is classically unstable, curvature stabilization must not be neglected, and the small island width stabilization effect from helical ion polarization currents is stronger than was previously thought [6]. The consequences of updated assumptions in ITER modeling of the minimum well-aligned ECCD power needed are all-in-all favorable (and well-within the ITER 24 gyrotron capability) when all effects are included. However, a “wild card” may be broadening of the localized

  20. A stopping criterion to halt iterations at the Richardson-Lucy deconvolution of radiographic images

    Science.gov (United States)

    Almeida, G. L.; Silvani, M. I.; Souza, E. S.; Lopes, R. T.

    2015-07-01

    Radiographic images, as any experimentally acquired ones, are affected by spoiling agents which degrade their final quality. The degradation caused by agents of systematic character, can be reduced by some kind of treatment such as an iterative deconvolution. This approach requires two parameters, namely the system resolution and the best number of iterations in order to achieve the best final image. This work proposes a novel procedure to estimate the best number of iterations, which replaces the cumbersome visual inspection by a comparison of numbers. These numbers are deduced from the image histograms, taking into account the global difference G between them for two subsequent iterations. The developed algorithm, including a Richardson-Lucy deconvolution procedure has been embodied into a Fortran program capable to plot the 1st derivative of G as the processing progresses and to stop it automatically when this derivative - within the data dispersion - reaches zero. The radiograph of a specially chosen object acquired with thermal neutrons from the Argonauta research reactor at Institutode Engenharia Nuclear - CNEN, Rio de Janeiro, Brazil, have undergone this treatment with fair results.

  1. Numerical study of a three-dimensional chemical vapor deposition reactor with detailed chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Ern, A. [CERMICS-ENPC, Noisy-le-Grand (France); Giovangigli, V. [Encole Polytechnique, Palaiseau (France); Smooke, M.D. [Yale Univ., New Haven, CT (United States)

    1996-06-01

    A numerical model of a three-dimensional, horizontal channel, chemical vapor deposition reactor is presented in order to study gallium arenide growth from trimethylgallium and arsine source reactants. Fluid flow and temperature predictions inside the reactor are obtained using the vorticity-velocity form of the three-dimensional, study-state Navier-Stokes equations coupled with a detailed energy balance equation inside the reactor and on its walls. Detailed gas phase and surface chemistry mechanisms are used to predict the chemical species profiles inside the reactor, the growth rate distribution on the substrate, and the level of carbon incorporation into the grown layer. The species diffusion velocities are written using the recent theory of iterative transport algorithms and account for both thermal diffusion and multicomponent diffusion processes. The influence of susceptor temperature and inlet composition on growth rate and carbon incorporation is found to agree well with previous numerical and experimental work. 46 refs., 15 figs., 3 tabs.

  2. Geometric feasibility of flexible cask transportation system for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Lima, P.; Ribeiro, M.I.; Aparicio, P. [Instituto Superior Tecnico-Instituto de Sistemas e Robotica, Lisboa (Portugal)

    1998-07-01

    One of the remote operations that has to be carried out in the International Thermonuclear Experimental Reactor (ITER) is the transportation of sealed casks between the various ports of the Tokamak Building (TB) and the Hot Cell Building (HCB). The casks may contain different in-vessel components (e.g. blanket modules, divertors) and are designed for a maximum load of about 80 ton. To improve the safety and flexibility of ITER Remote Handling (RH) transport vehicles, the cask is not motorized by itself, but instead, a motorized platform carrying the cask was proposed. This paper addresses the geometric feasibility of the flexible cask transportation system, taking into account the vehicle kinematics. The feasibility issues studied include planning smooth paths to increase safety, the discussion of building constraints by the evaluation of the vehicle spanned areas when following a planned path, and the analysis of the clearance required to remove the platform from underneath the cask at different possible failure locations. Simulation results are presented for the recommended trajectory, the spanned area and the rescue manoeuvres at critical locations along the path. (authors)

  3. Experimental study of contamination and cleaning of in-vessel mirrors for ITER optical diagnostics on T-10 and QSPA-T facilities

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, I., E-mail: igor_arkhipov_54@mail.ru [A.N. Frumkin Institute of Physical Chemistry and Electrochemistry Russian Academy of Sciences, Leninsky Avenue 31, Moscow 119991 (Russian Federation); Klimov, N. [SRC RF TRINITI, Pushkovykh Street 12, Troitsk, Moscow 142190 (Russian Federation); Svechnikov, N.; Grashin, S.; Vukolov, K.; Budaev, V.; Maslakov, K.; Stankevich, V. [NRC ‘Kurchatov Institute’, Kurchatov Square 1, Moscow 123182 (Russian Federation); Zhitlukhin, A.; Kovalenko, D.; Podkovyrov, V. [SRC RF TRINITI, Pushkovykh Street 12, Troitsk, Moscow 142190 (Russian Federation)

    2013-07-15

    ITER-relevant hydrocarbon films deposited under controlled conditions in the SOL of T-10 tokamak during working and cleaning discharges and by using QSPA-T plasma gun at plasma shots onto W and graphite targets were investigated by XPS, XAES and ellipsometry. XPS survey, valence band and C1s, O1s core level spectra showed differences in chemical composition and electronic structure as between T-10 and QSPA-T films and between T-10 films obtained in working and cleaning discharges. Hydrogen content evaluation in the films was made independently by analysis of XPS valence band edge position and ellipsometry data. The Csp{sup 2} carbon content was determined from XAES CKLL spectra. Classification of T-10 and QSPA-T films was undertaken using ternary phase diagram (H, Csp{sup 2} and Csp{sup 3}). It was shown that all studied hydrocarbon films are positioned between diamond- and polymer-like a-C:H films. The observed difference between XPS and ellipsometry evaluations of hydrogen content may be connected with strong structural non-uniformity of the films along thickness.

  4. Compact fusion reactors

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

  5. Autothermal reforming of methane with integrated CO2 capture in a novel fluidized bed membrane reactor. Part 1: experimental demonstration

    NARCIS (Netherlands)

    Gallucci, F.; van Sint Annaland, M.; Kuipers, J.A.M.

    2008-01-01

    Two fluidized bed membrane reactor concepts for hydrogen production via autothermal reforming of methane with integrated CO2 capture are proposed. Ultra-pure hydrogen is obtained via hydrogen perm-selective Pd-based membranes, while the required reaction energy is supplied by oxidizing part of the

  6. Experimental study of a cocurrent upflow packed bed bubble column reactor: pressure drop, holdup and interfacial area

    NARCIS (Netherlands)

    Molga, E.J.; Westerterp, K.R.

    1997-01-01

    Gas¿liquid interfacial areas have been determined by means of chemically enhanced absorption of CO2 into DEA in a packed bed bubble column reactor with an inner diameter of 156 mm. The influence of the gas velocity and particle diameter on the interfacial areas, pressure drops and liquid holdups has

  7. Fluidised bed membrane reactor for ultrapure hydrogen production via methane steam reforming: Experimental demonstration and model validation

    NARCIS (Netherlands)

    Patil, C.S.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Hydrogen is emerging as a future alternative for mobile and stationary energy carriers in addition to its use in chemical and petrochemical applications. A novel multifunctional reactor concept has been developed for the production of ultrapure hydrogen View the MathML source from light hydrocarbons

  8. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  9. Nuclear Analyses of Indian LLCB Test Blanket System in ITER

    Science.gov (United States)

    Swami, H. L.; Shaw, A. K.; Danani, C.; Chaudhuri, Paritosh

    2017-04-01

    Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2, to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System design & development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radioactive waste management, equipment maintenance & replacement strategies and nuclear safety. The nuclear behaviour of LLCB test blanket module in ITER is predicated in terms of nuclear responses such as tritium production, nuclear heating, neutron fluxes and radiation damages. Radiation shielding capability of LLCB TBS inside and outside bio-shield was also assessed to fulfill ITER shielding requirements. In order to supports the rad-waste and safety assessment, nuclear activation analyses were carried out and radioactivity data were generated for LLCB TBS components. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.l). The paper describes a comprehensive nuclear performance of LLCB TBS in ITER.

  10. Test results of the first 50 kA NbTi full size sample for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ciazynski, D.; Zani, L. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Ciotti, M.; Gislon, P.; Spadoni, M. [Association Euratom-ENEA sulla Fusione Centro Ricerche Energia Frascati (Italy); Huber, S.; Stepanov, B. [Association Euratom-Confederation Suisse, EPFL, CRPP-TF, Villigen (Switzerland); Karlemo, B. [European Fusion Development Agreement - Close Support Unit (EFDA/CSU), Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2003-07-01

    Within the framework of the research studies for the International Thermonuclear Experimental Reactor (ITER) project, the first full size NbTi conductor sample was fabricated in industry and tested in the SULTAN facility (Villigen, Switzerland). This sample (PF-FSJS), which is relevant to the Poloidal Field coils of ITER, is composed of two parallel straight bars of conductor, connected at bottom through a joint designed according to the Cea twin-box concept. The two conductor legs are identical except for the use of different strands: a nickel plated NbTi strand with a pure copper matrix in one leg, and a bare NbTi strand with copper matrix and internal CuNi barrier in the other leg. The two conductors and the joint were extensively tested regarding DC (direct current) and AC (alternative current) properties. This paper reports on the tests results and analysis, stressing the differences between the two conductor legs and discussing the impact of the test results on the ITER design criteria for conductor and joint. While joint DC resistance, conductors and joint AC losses, fulfilled the ITER requirements, neither conductor could reach its current sharing temperature at relevant ITER currents, due to instabilities. Although the drop in temperature is slight for the CuNi strand cable, it is more significant for the Ni plated strand cable. (authors)

  11. Recommendations on the Nature and Level of U.S. Participation in the International Thermonuclear Experimental Reactor Extension of the Experimental Reactor Extension of the Engineering Design Activities. Panel Report To Fusion Energy Sciences Advisory Committee (FESAC)

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1998-01-31

    The DOE Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to "address the topic of U. S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction." (Attachment 1: DOE Charge, September 1996). Given that there is expected to be a transition period of three to five years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to "include the U.S. role in an interim period between the EDA and construction." (Attachment 2: DOE Expanded Charge, May 1997). This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirety. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.

  12. Report made on behalf of the commission of foreign affairs about the project of law, adopted by the Senate, giving permission to the approval of the agreement between the French government and the international organization for thermonuclear fusion energy ITER, relative to the head office of ITER organization and to the privileges and immunities of ITER organization in the French territory; Rapport fait au nom de la commission des affaires etrangeres sur le projet de loi, adopte par le senat, autorisant l'approbation de l'accord entre le Gouvernement de la Republique francaise et l'Organisation internationale ITER pour l'energie de fusion relatif au siege de l'Organisation ITER et aux privileges et immunites de l'Organisation ITER sur le territoire francais

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-02-15

    The will of building up an international thermonuclear experimental reactor (ITER) gathers since several years the European community of atomic energy (Euratom), Japan, the USA, and Russia, next followed by China, South Korea and, since 2005, by India. The agreement signed in Paris between these seven parties on November 21, 2006 entrusted the international organization ITER with the realization of this project. The implications of the ITER project are enormous both in their scientific and in their economical aspects. France has a particular position in this project since the head office of ITER organisation is sited at Saint-Paul-lez-Durance and the tokamak will be built at Cadarache. Therefore, an agreement has been signed between ITER organization and the French government. The approval of this agreement is the object of this project of law. The document presents first the principle, challenge and stakes of the ITER project and the long negotiations that have led to chose France for the setting up of the research facility. Then, it presents the agreement of November 7, 2007, which fixes the general framework of the conditions necessary to the achievement of the project. The agreement comprises the classical rules, privileges and immunities of any international organization sited on the French ground, plus some more specific dispositions in particular in the domain of management of the wastes generated by the facility. (J.S.)

  13. Experimental determination of spectral indices by scanning of fuel rod in the IPEN/MB-01 reactor; Determinacao experimental de indices espectrais por varredura gama de vareta combustivel no reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Fanaro, Leda Cristina Cabelo Bernardes

    2009-07-01

    In this work, the spectral indexes 28{rho}{sup *} and 25{delta}{sup *}, and gamma efficiency factor in the IPEN/MB-01 reactor were determined experimentally employing a rod scanning technique. In the case of 28{rho}{sup *}, this method has the advantage of eliminating most of the correction factors derived from the calculations. Only the fission yield factor and the relative fission rate in the {sup 235}U remain in the determination of the 25{delta}{sup *}. The experiments were performed with different thicknesses of cadmium sleeves: 0.55 mm, 1.10 mm and 2.20 mm. The final experimental uncertainty achieved in the experiment, less that 1 %, and the excellent geometrical and material data characterization of the IPEN/MB-01 reactor allow to use the results as benchmark for validate calculation methods and related nuclear data libraries. The comparison between calculated values and experimental values was performed by employing the MCNP-5 code and the nuclear data libraries: ENDF/B-VI.8, ENDF/B-VII.0, JENDL-3.3 and JEFF-3.1. The results demonstrate that the difference among libraries is very small. Also, the comparison between calculated values and experimental values shows that there has been considerable progress in the recent nuclear data libraries. The best result is obtained with ENDF/B-VII.0 nuclear data library, and the highest discrepancy was obtained with JEFF-3.1 and JENDL-3.3 nuclear data libraries. (author)

  14. Energy efficient electrocoagulation using a new flow column reactor to remove nitrate from drinking water - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Pedrola, Montserrat Ortoneda; Phipps, David

    2017-07-01

    In this investigation, a new bench-scale electrocoagulation reactor (FCER) has been applied for drinking water denitrification. FCER utilises the concepts of flow column to mix and aerate the water. The water being treated flows through the perforated aluminium disks electrodes, thereby efficiently mixing and aerating the water. As a result, FCER reduces the need for external stirring and aerating devices, which until now have been widely used in the electrocoagulation reactors. Therefore, FCER could be a promising cost-effective alternative to the traditional lab-scale EC reactors. A comprehensive study has been commenced to investigate the performance of the new reactor. This includes the application of FCER to remove nitrate from drinking water. Estimation of the produced amount of H2 gas and the yieldable energy from it, an estimation of its preliminary operating cost, and a SEM (scanning electron microscope) investigation of the influence of the EC process on the morphology of the surface of electrodes. Additionally, an empirical model was developed to reproduce the nitrate removal performance of the FCER. The results obtained indicated that the FCER reduced the nitrate concentration from 100 to 15 mg/L (World Health Organization limitations for infants) after 55 min of electrolysing at initial pH of 7, GBE of 5 mm, CD of 2 mA/cm2, and at operating cost of 0.455 US $/m3. Additionally, it was found that FCER emits H2 gas enough to generate a power of 1.36 kW/m3. Statistically, the relationship between the operating parameters and nitrate removal could be modelled with R2 of 0.848. The obtained SEM images showed a large number dents on anode's surface due to the production of aluminium hydroxides. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

  15. Fusion Technology for ITER, the ITER Project. Further Development Towards a DEMO Fusion Power Plant (3/4)

    CERN Document Server

    CERN. Geneva

    2011-01-01

    This is the second half of a lecture series on fusion and will concentrate on fusion technology. The early phase of fusion development was concentrated on physics. However, during the 1980s it was realized that if one wanted to enter the area of fusion reactor plasmas, even in an experimental machine, a significant advance in fusion technologies would be needed. After several conceptual studies of reactor class fusion devices in the 1980s the engineering design phase of ITER started in earnest during the 1990s. The design team was in the beginning confronted with many challenges in the fusion technology area as well as in physics for which no readily available solution existed and in a few cases it was thought that solutions may be impossible to find. However, after the initial 3 years of intensive design and R&D work in an international framework utilizing basic fusion technology R&D from the previous decade it became clear that for all problems a conceptual solution could be found and further devel...

  16. Potential for use of high-temperature superconductors in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hull, J.R.

    1991-01-01

    The present rate of development of high-temperature superconductors (HTSs) is sufficiently rapid that there may be opportunities for their use in contemporary fusion devices such as the International Thermonuclear Experimental Reactor (ITER). The most likely 1application is for delivering power to the superconducting magnets, especially in substituting for the current leads between the temperatures of 4 K and 77K. A second possible application of HTSs is as a liquid-nitrogen-cooled power bus, connecting the power supplies to the magnets, thus reducing the ohmic heating losses over these relatively long cables. A third potential application of HTSs is as an inner high-field winding of the toroidal field coils that would operate at {approx}20 K. While the use of higher temperature magnets offers significant advantages to the reactor system, it is unlikely that tested conductors of this type will be available within the ITER time frame. 23 refs., 2 figs.

  17. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  18. Determination of the Clean Air Delivery Rate (CADR) of Photocatalytic Oxidation (PCO) Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results.

    Science.gov (United States)

    Héquet, Valérie; Batault, Frédéric; Raillard, Cécile; Thévenet, Frédéric; Le Coq, Laurence; Dumont, Éric

    2017-03-06

    The performances of a laboratory PhotoCatalytic Oxidation (PCO) device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m³ and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10-3 m³. The PCO device was composed of a pleated photocatalytic filter (1100 cm²) and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR) of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i) light irradiance I from 0.10 to 2.0 mW·cm-2; (ii) air velocity v from 0.2 to 1.9 m·s-1; and (iii) initial toluene concentration C₀ (200, 600, 1000 and 4700 ppbv). The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I) were fulfilled. The CADR values, ranging from 0.35 to 3.95 m³·h-1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I), the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  19. Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results

    Directory of Open Access Journals (Sweden)

    Valérie Héquet

    2017-03-01

    Full Text Available The performances of a laboratory PhotoCatalytic Oxidation (PCO device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m3 and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10−3 m3. The PCO device was composed of a pleated photocatalytic filter (1100 cm2 and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i light irradiance I from 0.10 to 2.0 mW·cm−2; (ii air velocity v from 0.2 to 1.9 m·s−1; and (iii initial toluene concentration C0 (200, 600, 1000 and 4700 ppbv. The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I were fulfilled. The CADR values, ranging from 0.35 to 3.95 m3·h−1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I, the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  20. Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop applied for a fusion reactor Flibe blanket

    Energy Technology Data Exchange (ETDEWEB)

    Toda, Saburo; Chiba, Shinya E-mail: schiba@karma.qse.tohoku.ac.jp; Yuki, Kazuhisa; Omae, Masahiro; Sagara, Akio

    2002-12-01

    Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop (MSL) is described in this paper. The MSL was designed to be able to use Flibe as a coolant, however, a simulant, heat transfer salt (HTS) has to be used alternatively since Flibe is difficult to operate under avoiding a biohazard of Be. Experiment on heat-transfer enhancement, that is required for applying to cool the high heat flux components of fusion reactors, is ongoing. Preliminary experimental results showed that an internal structure of a mixing chamber in the MSL was important to obtain accurate bulk temperatures under severe thermal conditions. For operating the loop, careful handling are needed to proceed how to melt the salt and to circulate it in starting the operation of the MSL. It is concluded that several improvements proposed from the present experiences should be applied for the future Flibe operation.

  1. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  2. Neutronics qualification of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program - Transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

    Energy Technology Data Exchange (ETDEWEB)

    Leray, O.; Hudelot, J. P. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Antony, M. [DEN Experimental Physics Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Doederlein, C. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Santamarina, A. [DEN Reactor and Cycle Physics Service, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Bernard, D. [DEN Physical Studies Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France); Vaglio-Gaudard, C. [DEN Nuclear Projects Laboratory, CEA Cadarache, F-13108 St Paul-Lez-Durance (France)

    2011-07-01

    The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in {sup 235}U) fuels (U{sub 3}Si{sub 2} for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has been performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl{sub x}/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm{sup 3}. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% {sup 235}U) by the Minerve

  3. An iterative method for determination of a minimal eigenvalue

    DEFF Research Database (Denmark)

    Kristiansen, G.K.

    1968-01-01

    Kristiansen (1963) has discussed the convergence of a group of iterative methods (denoted the Equipoise methods) for the solution of reactor criticality problems. The main result was that even though the methods are said to work satisfactorily in all practical cases, examples of divergence can be...

  4. Development of Flow-Through Polymeric Membrane Reactor for Liquid Phase Reactions: Experimental Investigation and Mathematical Modeling

    Directory of Open Access Journals (Sweden)

    Endalkachew Chanie Mengistie

    2017-01-01

    Full Text Available Incorporating metal nanoparticles into polymer membranes can endow the membranes with additional functions. This work explores the development of catalytic polymer membrane through synthesis of palladium nanoparticles based on the approaches of intermatrix synthesis (IMS inside surface functionalized polyethersulfone (PES membrane and its application to liquid phase reactions. Flat sheet PES membranes have been successfully modified via UV-induced graft polymerization of acrylic acid monomer. Palladium nanoparticles have been synthesized by chemical reduction of palladium precursor loaded on surface modified membranes, an approach to the design of membranes modified with nanomaterials. The catalytic performances of the nanoparticle incorporated membranes have been evaluated by the liquid phase reduction of p-nitrophenol using NaBH4 as a reductant in flow-through membrane reactor configuration. The nanocomposite membranes containing palladium nanoparticles were catalytically efficient in achieving a nearly 100% conversion and the conversion was found to be dependent on the flux, amount of catalyst, and initial concentration of nitrophenol. The proposed mathematical model equation represents satisfactorily the reaction and transport phenomena in flow-through catalytic membrane reactor.

  5. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  6. Robust iterative methods

    Energy Technology Data Exchange (ETDEWEB)

    Saadd, Y.

    1994-12-31

    In spite of the tremendous progress achieved in recent years in the general area of iterative solution techniques, there are still a few obstacles to the acceptance of iterative methods in a number of applications. These applications give rise to very indefinite or highly ill-conditioned non Hermitian matrices. Trying to solve these systems with the simple-minded standard preconditioned Krylov subspace methods can be a frustrating experience. With the mathematical and physical models becoming more sophisticated, the typical linear systems which we encounter today are far more difficult to solve than those of just a few years ago. This trend is likely to accentuate. This workshop will discuss (1) these applications and the types of problems that they give rise to; and (2) recent progress in solving these problems with iterative methods. The workshop will end with a hopefully stimulating panel discussion with the speakers.

  7. Experimental and modeling study of the effect of CO and H2 on the urea DeNO(x) process in a 150kW laboratory reactor.

    Science.gov (United States)

    Javed, M Tayyeb; Nimmo, W; Gibbs, B M

    2008-01-01

    An experimental and modeling investigation has been performed to study the effect of process additives, H2 and CO on NO(x) removal from flue gases by a selective non-catalytic reduction process using urea as a reducing agent. Experiments were performed with a flow reactor in which flue gas was generated by the combustion of propane in air at 3% excess oxygen and the desired levels of initial NO(x) (500ppm) were achieved by doping the flame with ammonia. Experiments were performed throughout the temperature range of interest, i.e. from 850 to 1200 degrees C for investigation of the effects of the process additives on the performance of aqueous urea DeNO(x). Subsequently, computational kinetic modeling with SENKIN code was performed to analyze the performance of urea providing a direct comparison of modeling prediction with experimental measurements. With CO addition, a downwards shift of 215 degrees C in the peak reduction temperature from 1125 to 910 degrees C was observed during the experimentation while the kinetic modeling suggests it to be 150 degrees C, i.e. from 1020 to 870 degrees C. The addition of H2 impairs the peak NO(x) reduction but suggests a low temperature application of the process. A downward shift of 250 degrees C in the peak reduction temperature, from 1020 to 770 degrees C, was observed during kinetic modeling studies. The kinetic modeling shows a good qualitative agreement with the experimental observations and reveals additional information about the process.

  8. Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-02-01

    This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

  9. Development of pipe welding, cutting and inspection tools for the ITER blanket

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Taguchi, Kou; Takiguchi, Yuji; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-07-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor (ITER), an internal access welding/cutting of blanket cooling pipe with bend sections is inevitably required because of spatial constraint due to nuclear shield and available port opening space. For this purpose, internal access pipe welding/cutting/inspection tools for manifolds and branch pipes are being developed according to the agreement of the ITER R and D task (T329). A design concept of welding/cutting processing head with a flexible optical fiber has been developed and the basic feasibility studies on welding, cutting and rewelding are performed using stainless steel plate (SS316L). In the same way, a design concept of inspection head with a non-destructive inspection probe (including a leak-testing probe) has been developed and the basic characteristic tests are performed using welded stainless steel pipes. In this report, the details of welding/cutting/inspection heads for manifolds and branch pipes are described, together with the basic experiment results relating to the welding/cutting and inspection. In addition, details of a composite type optical fiber, which can transmit both the high-power YAG laser and visible rays, is described. (author)

  10. Sequential Aeration of Membrane-Aerated Biofilm Reactors for High-Rate Autotrophic Nitrogen Removal: Experimental Demonstration

    DEFF Research Database (Denmark)

    Pellicer i Nàcher, Carles; Sun, Sheng-Peng; Lackner, Susanne

    2010-01-01

    One-stage autotrophic nitrogen (N) removal, requiring the simultaneous activity of aerobic and anaerobic ammonium oxidizing bacteria (AOB and AnAOB), can be obtained in spatially redox-stratified biofilms. However, previous experience with Membrane-Aerated Biofilm Reactors (MABRs) has revealed......S rRNA gene confirmed that sequential aeration, even at elevated average O2 loads, stimulated the abundance of AnAOB and AOB and prevented the increase in NOB. Nitrous oxide (N2O) emissions were 100-fold lower compared to other anaerobic ammonium oxidation (Anammox)-nitritation systems. Hence...... a difficulty in reducing the abundance and activity of nitrite oxidizing bacteria (NOB), which drastically lowers process efficiency. Here we show how sequential aeration is an effective strategy to attain autotrophic N removal in MABRs: Two separate MABRs, which displayed limited or no N removal under...

  11. Pilot Experimentation with Complete Mixing Anoxic Reactors to Improve Sewage Denitrification in Treatment Plants in Small Communities

    Directory of Open Access Journals (Sweden)

    Massimo Raboni

    2013-12-01

    Full Text Available This paper reports the results of two sewage treatment tests in a community of 15,000 inhabitants. The sewage treatment plant is subject to strong fluctuations in load (BOD5, COD, TKN, and in particular in the BOD5/TKN ratio. These fluctuations adversely affect the biological denitrification, as demonstrated by many pilot and real-scale plants. The plants we tested were subjected to two treatment types: anoxic-aerobic and simultaneous denitrification. Both processes are designed for complete mixing conditions in the reactors in order to level the fluctuations in the load and thus improve the denitrification efficiency. The results prove that an average denitrification efficiency of up to 80% can be achieved with the sludge loading close to 0.1 kg BOD5 (d∙kgMLVSS−1. The effect of the sludge loading and dissolved oxygen on the denitrification efficiency is highlighted.

  12. Defluoridation of drinking water using a new flow column-electrocoagulation reactor (FCER) - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Ortoneda Pedrola, Montserrat; Phipps, David

    2017-07-15

    A new batch, flow column electrocoagulation reactor (FCER) that utilises a perforated plate flow column as a mixer has been used to remove fluoride from drinking water. A comprehensive study has been carried out to assess its performance. The efficiency of fluoride removal (R%) as a function of key operational parameters such as initial pH, detention time (t), current density (CD), inter-electrode distance (ID) and initial concentration (C 0 ) has been examined and an empirical model has been developed. A scanning electron microscopy (SEM) investigation of the influence of the EC process on morphology of the surface of the aluminium electrodes, showed the erosion caused by aluminium loss. A preliminary estimation of the reactor's operating cost is suggested, allowing for the energy from recycling of hydrogen gas hydrogen gas produced amount. The results obtained showed that 98% of fluoride was removed within 25 min of electrolysis at pH of 6, ID of 5 mm, and CD of 2 mA/cm 2 . The general relationship between fluoride removal and operating parameters could be described by a linear model with R 2 of 0.823. The contribution of the operating parameters to the suggested model followed the order: t > CD > C 0  > ID > pH. The SEM images obtained showed that, after the EC process, the surface of the anodes, became non-uniform with a large number of irregularities due to the generation of aluminium hydroxides. It is suggested that these do not materially affect the performance. A provisional estimate of the operating cost was 0.379 US $/m 3 . Additionally, it has been found that 0.6 kW/m 3 is potentially recoverable from the H 2 gas. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

  13. Development of BWR regional stability experimental facility SIRIUS-F, which simulates thermal-hydraulics-neutronics coupling in reactor core, and stability evaluation of ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Masahiro Furuya; Fumio Inada [Central Research Institute of Electric Power Industry (CRIEPI) 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Takanori Fukahori [Global Nuclear Fuel Japan (GNF-J) 2-3-1 Uchikawa, Yokosuka, Kanagawa 239-0836 (Japan); Shinya Mizokami [Tokyo Electric Power Company (TEPCO) 1-1-3 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)

    2005-07-01

    Full text of publication follows: The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement. (authors)

  14. Experimental study and nuclear model calculations on the 192Os(p,n)192Ir reaction: Comparison of reactor and cyclotron production of the therapeutic radionuclide 192Ir.

    Science.gov (United States)

    Hilgers, K; Sudár, S; Qaim, S M

    2005-07-01

    In a search for an alternative route of production of the important therapeutic radionuclide (192)Ir (T(1/2)=78.83 d), the excitation function of the reaction (192)Os(p,n)(192)Ir was investigated from its threshold up to 20 MeV. Thin samples of enriched (192)Os were obtained by electrodeposition on Ni, and the conventional stacked-foil technique was used for cross section measurements. The experimental data were compared with the results of theoretical calculations using the codes EMPIRE-II and ALICE-IPPE. Good agreement was found with EMPIRE-II, but slightly less with the ALICE-IPPE calculations. The theoretical thick target yield of (192)Ir over the energy range E(p)=16-->8 MeV amounts to only 0.16MBq/muA.h. A comparison of the reactor and cyclotron production methods is given. In terms of yield and radionuclidic purity of (192)Ir the reactor method appears to be superior; the only advantage of the cyclotron method could be the higher specific activity of the product.

  15. Experimental study and nuclear model calculations on the $^{192}Os (p, n)^{192}$Ir reaction Comparison of reactor and cyclotron production of the therapeutic radionuclide $^{192}$Ir

    CERN Document Server

    Hilgers, K; Sudar, S; 10.1016/j.apradiso.2004.12.010

    2005-01-01

    In a search for an alternative route of production of the important therapeutic radionuclide /sup 192/Ir (T/sub 1/2/=78.83 d), the excitation function of the reaction /sup 192/Os(p, n)/sup 192/Ir was investigated from its threshold up to 20MeV. Thin samples of enriched /sup 192/Os were obtained by electrodeposition on Ni, and the conventional stacked-foil technique was used for cross section measurements. The experimental data were compared with the results of theoretical calculations using the codes EMPIRE-II and ALICE-IPPE. Good agreement was found with EMPIRE-II, but slightly less with the ALICE-IPPE calculations. The theoretical thick target yield of /sup 192/Ir over the energy range E/sub p/=16 to 8MeV amounts to only 0.16MBq/ mu A.h. A comparison of the reactor and cyclotron production methods is given. In terms of yield and radionuclidic purity of /sup 192/Ir the reactor method appears to be superior; the only advantage of the cyclotron method could be the higher specific activity of the product.

  16. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  17. Activation of water coolant in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, V.; Santoro, R.T.; Lida, H.; Parker, R.; Janeschitz, G.; Plenteda, R. [ITER Joint Central Team Garching, Muenchen (Germany)

    1998-07-01

    Water as been selected as the main coolant for the ITER blanket and vacuum vessel. Following exposure to DT neutrons, water becomes a source of high energy {sup 16}N-decay photons and energetic ({approx}0.9 MeV) {sup 17}N decay neutrons outside the reactor that lead to shielding problems during both reactor operation and after shutdown. As a result of comprehensive neutronic and hydraulic analyses, corresponding design measures were developed to diminish these effects. The use of a correlation between the {sup 16}N-production rate and the 14.1-MeV neutron flux in flowing water was recommended for determining fusion power by measuring water decay photons behind the radiation shield. (authors)

  18. Calculations of Energy Losses due to Atomic Processes in Tokamaks with Applications to the ITER Divertor

    CERN Document Server

    Post, D; Clark, R E H; Putvinskaya, N

    1995-01-01

    Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) 1 . Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations are the result of a complex interplay of many processes. In order to identify the key features of these experiments and calculations and the relative role of the primary atomic processes, simple quasi-analytic models and the latest atomic physics rate coefficients and cross sections have been used to assess the relative roles of central radiation losses through bremsstrahlung, impurity radiation losses from the plasma edge, charge exchange and hydrogen radiation losses f...

  19. Experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors; Experimentelle und numerische Untersuchung von Gas/Liquid-Phasengrenzflaechen als Referenzwert fuer die hydrostatische Fuellstandsmessung in Siedewasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, Stephan

    2013-12-17

    The experimental and numerical investigation of gas/liquid phase boundaries representing the reference level for hydrostatic level measurements in boiling water reactors is considered as relevant for reactor safety research. The experiments allow a quantification of the transition processes in hydrostatic level measurement devices that were up to now only assessed by phenomenological descriptions. Experimental studies covered the topology and stability of water/vapor phase boundaries and the numerical description using CFD codes, including modeling of the surface topology and modeling of the heat and mass transport.

  20. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. Manufacturing and maintenance technologies developed for a thick-wall structure of the ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M. E-mail: onozukm@itereu.de; Alfile, J.P.; Aubert, Ph.; Dagenais, J.-F.; Grebennikov, D.; Ioki, K.; Jones, L.; Koizumi, K.; Krylov, V.; Maslakowski, J.; Nakahira, M.; Nelson, B.; Punshon, C.; Roy, O.; Schreck, G

    2001-09-01

    Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems have been carried out for on-site manufacturing and maintenance of the thick-wall structure of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV). Conventional techniques, including tungsten inert gas welding, plasma cutting, and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated, including reduced-pressure electron-beam and multi-pass neodymium-doped yttrium aluminum garnet (NdYAG) laser welding, NdYAG laser cutting, and electro-magnetic acoustic transducer inspection, to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.

  2. Design assessment for manufacturability of supporting structures for IWS and ELM coil of ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yu-Gyeong, E-mail: aspirany@hhi.co.kr [Hyundai Heavy Industries Co., Ltd., Dong-gu, Ulsan (Korea, Republic of); Hong, Yun-Seok; Kwon, Tae-Hoon; Jung, Yung-Jin [Hyundai Heavy Industries Co., Ltd., Dong-gu, Ulsan (Korea, Republic of); Ahn, Hee-Jae; Kim, Byung-Chul; Lee, Young-Ju [National Fusion Research Institute, 113 Gwahangno, Yuseong-gu, Daejeon-si 305-333 (Korea, Republic of)

    2013-10-15

    Highlights: ► A “bridge type” welding method was proposed for IWS rib and ELM coil supports. ► This method could reduce the welding distortion and increase the manufacturability. ► Structural integrity of proposed design was assessed for both P and S type damages. ► Results show that the proposed designs meet the design criteria of RCC-MR. -- Abstract: A vacuum vessel is one of the core facilities of ITER (International Thermonuclear Experimental Reactor) and basically all-welded structure. Korea is responsible for the procurement of sector 1 and 6 of the main vessel. Accordingly, the design review for the fabrication is in progress by ITER Korea and Hyundai Heavy Industries. Due to anticipated manufacturing problems such as the welding distortion, the design of some components of main vessel, IWS (In-Wall Shield) supporting rib and ELM (Edge Localized Mode) coil support, needs to be modified. To release the risk of welding distortion, the welding method called “bridge type” is suggested and the shape of weld joint is adjusted to secure the manufacturability of the issued components. The elastic and limit analyses with fatigue evaluation have been performed under the most critical loading condition to verify the structural integrity of modified design. Analysis results show that the proposed designs meet the design criteria of RCC-MR. The design deviation requests have been submitted to ITER Organization and ANB (Agreed Notified Body) for approval and their verification is currently in progress.

  3. A design study of hydrogen isotope separation system for ITER-FEAT

    Energy Technology Data Exchange (ETDEWEB)

    Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    Preliminary design study of the hydrogen isotope separation system (ISS) for the fuel cycle of the ITER-FEAT, a fusion experimental reactor, was carried out based on the substantial reduction of hydrogen flow to the ISS resulting from the design study for scale reduction of the formerly-designed ITER. Three feed streams (plasma exhaust gas stream, streams from the water detritiation system and that from the neutral beam injectors) are fed to the ISS, and three product streams (high purity tritium gas, high purity deuterium gas and hydrogen gas) are made in it by the method of cryogenic distillation. In this study, an original four-column cascade was proposed to the ISS cryogenic distillation column system considering simplification and the operation scenario of the ITER-FEAT. Substantial reduction of tritium inventory in the ISS was found to be possible in the progress of investigation concerning of the corresponding flow rate of tritium product stream (T>90 %) for pellet injector which depends upon the operation condition. And it was found that tritium concentration in the released hydrogen stream into environment from the ISS could easily fluctuate with current design of column arrangement due to the small disturbance in mass flow balance in the ISS. To solve this problem, two-column system for treatment of this flow was proposed. (author)

  4. Robot vision system R and D for ITER blanket remote-handling system

    Energy Technology Data Exchange (ETDEWEB)

    Maruyama, Takahito, E-mail: maruyama.takahito@jaea.go.jp [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Aburadani, Atsushi; Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki-ken 311-0193 (Japan); Tesini, Alessandro [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France)

    2014-10-15

    For regular maintenance of the International Thermonuclear Experimental Reactor (ITER), a system called the ITER blanket remote-handling system is necessary to remotely handle the blanket modules because of the high levels of gamma radiation. Modules will be handled by robotic power manipulators and they must have a non-contact-sensing system for installing and grasping to avoid contact with other modules. A robot vision system that uses cameras was adopted for this non-contact-sensing system. Experiments for grasping modules were carried out in a dark room to simulate the environment inside the vacuum vessel and the robot vision system's measurement errors were studied. As a result, the accuracy of the manipulator's movements was within 2.01 mm and 0.31°, which satisfies the system requirements. Therefore, it was concluded that this robot vision system is suitable for the non-contact-sensing system of the ITER blanket remote-handling system.

  5. Experimental and numerical investigation of coolant mixing in a model of reactor pressure vessel down-comer and in cold leg inlets

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2017-01-01

    Full Text Available Thermal fatigue and pressurized thermal shock phenomena are the main problems for the reactor pressure vessel and the T-junctions both of them depend on the mixing of the coolant. The mixing process, flow and temperature distribution has been investigated experimentally using particle image velocimetry, laser induced fluorescence, and simulated by CFD tools. The obtained results showed that the ratio of flow rate between the main pipe and the branch pipe has a big influence on the mixing process. The particle image velocimetry/planar laser-induced fluorescence measurements technologies proved to be suitable for the investigation of turbulent mixing in the complicated flow system: both velocity and temperature distribution are important parameters in the determination of thermal fatigue and pressurized thermal shock. Results of the applied these techniques showed that both of them can be used as a good provider for data base and to validate CFD results.

  6. CFD analysis of flow boiling in the ITER first wall

    Energy Technology Data Exchange (ETDEWEB)

    Domalapally, Phani, E-mail: phani.domalapally@polito.it [Dipartimento di Energetica, Politecnico, I-10129 Torino (Italy); Rizzo, Enrico [Institut fuer Technische Physik, Karlsruhe Institute of Technology, Karlsruhe (Germany); Savoldi Richard, Laura; Subba, Fabio; Zanino, Roberto [Dipartimento di Energetica, Politecnico, I-10129 Torino (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Two boiling models, Rohsenow model for nucleate boiling, switching to the Volume of Fluid for film boiling using STAR-CCM+ and Bergles-Rohsenow model, for which we developed a User Defined Function, implemented in FLUENT, are tested and compared on a flat-channel geometry. Black-Right-Pointing-Pointer Performance is checked in terms of accuracy and computational cost. Black-Right-Pointing-Pointer As a first approximation analysis Bergles-Rohsenow model may be a good choice, if more accurate information on thermal behavior or flow field is required, the other should be preferred. - Abstract: This paper compares two Computational Fluid Dynamic (CFD) approaches for the analysis of flow boiling inside the first wall (FW) of the International Thermonuclear Experimental Reactor (ITER): (1) the Rohsenow model for nucleate boiling, seamlessly switching to the Volume of Fluid (VOF) approach for film boiling, as available in the commercial CFD code STAR-CCM+, (2) the Bergles-Rohsenow (BR) model, for which we developed a User Defined Function (UDF), implemented in the commercial code FLUENT. The physics of both models is described, and the results with different inlet conditions and heating levels are compared with experimental results obtained at the Efremov Institute, Russia. The performance of both models is compared in terms of accuracy and computational cost.

  7. Metal Hall sensors for the new generation fusion reactors of DEMO scale

    Science.gov (United States)

    Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

    2017-11-01

    For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (fusion reactors of DEMO scale.

  8. Experimental Investigation of the Root Cause Mechanism and Effectiveness of Mitigating Actions for Axial Offset Anomaly in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Said Abdel-Khalik

    2005-07-02

    Axial offset anomaly (AOA) in pressurized water reactors refers to the presence of a significantly larger measured negative axial offset deviation than predicted by core design calculations. The neutron flux depression in the upper half of high-power rods experiencing significant subcooled boiling is believed to be caused by the concentration of boron species within the crud layer formed on the cladding surface. Recent investigations of the root-cause mechanism for AOA [1,2] suggest that boron build-up on the fuel is caused by precipitation of lithium metaborate (LiBO2) within the crud in regions of subcooled boiling. Indirect evidence in support of this hypothesis was inferred from operating experience at Callaway, where lithium return and hide-out were, respectively, observed following power reductions and power increases when AOA was present. However, direct evidence of lithium metaborate precipitation within the crud has, heretofore, not been shown because of its retrograde solubility. To this end, this investigation has been undertaken in order to directly verify or refute the proposed root-cause mechanism of AOA, and examine the effectiveness of possible mitigating actions to limit its impact in high power PWR cores.

  9. 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.

  10. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  11. First experimental results of particle re-suspension in a low pressure wind tunnel applied to the issue of dust in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rondeau, Anthony, E-mail: anthony.rondeau@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA, Gif-sur-Yvette 91192 (France); Merrison, Jonathan; Iversen, Jens Jacob [Department of Physics and Astronomy, Institute for Storage Ring Facilities, 8000 Aarhus C (Denmark); Peillon, Samuel; Sabroux, Jean-Christophe; Lemaitre, Pascal; Gensdarmes, François [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SCA, Gif-sur-Yvette 91192 (France); Chassefière, Eric [Laboratoire Géosciences Paris Sud (GEOPS), UMR 8148, Université Paris Sud, 91403 Orsay Cedex (France)

    2015-10-15

    Highlights: • The first experimental data of dust re-suspension performed by controlled airflow (in terms of airflow velocity and fluid density). • The effect of the surrounding pressure in the re-suspension mechanism. • The friction (shear velocity) reduction at low pressure in the transient regime (Knudsen number close to one). • The importance of the adhesion forces between particles in dust mobilization by clustering. • The fact that the existing re-suspension models of the relevant literature do not take into account this clustering phenomenon. - Abstract: During the normal operating condition of the future ITER tokamak, a massive production of dust in the toroidal vacuum vessel is expected. This dust, originating from the erosion of tungsten and beryllium internal walls of the torus by the plasma, would be mobilized to some extent during a loss of vacuum accident (LOVA). For safety reasons, it is essential to quantify the re-suspended dust fraction during such an event. Here, we provide preliminary experimental data of dust re-suspension obtained in the wind tunnel of the European Space Agency (ESA) at low pressures (300, 130 and 10 mbar). The experimentations were performed with multilayer deposits. We used two powders with a median diameter at 15.5 μm and 21.8 μm. A negative influence of the low pressure in the re-suspension mechanism is observed. For example, given a re-suspension fraction of 10%, increasing friction shear velocities are derived for decreasing absolute pressures: 300 mbar/0.66 m s{sup −1}; 130 mbar/1.08 m s{sup −1}; and 10 mbar/1.84 m s{sup −1}. In addition, we highlight the friction reduction for Kundsen numbers greater than 0.1 by an analysis of the airflow forces.

  12. Fluid flow separation in a reactor pressure vessel during an ECC injection. Single phase flow and two phase flow (air-water) experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Bichet; Alain Martin [EDF - Research and Development Division - Fluid Mechanics and Heat Transfert 6, quai Watier - B.P. 49 - 78401 Chatou CEDEX 01 (France); Frederic Beaud [EDF/ Industry - Basic Design Department., 12-14, Avenue Dutrievoz 69628 Villeurbanne CEDEX (France)

    2005-07-01

    Full text of publication follows: Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) integrity has been performed. A simplified analysis has shown that the most severe loading conditions are given by the small break loss of coolant accidents due to the pressurized injection of cold water (9 deg. C) into the cold leg and down comer of the RPV. During these transient scenarios, single or two-phase (uncovered cold leg) flows have been shown in the cold leg, depending on the crack size and RPV model (900 MWe or 1300 MWe). An experimental study has been carried out, on the one hand, to consolidate the numerical results obtained with CFD home code (Code-Saturne) which mainly showed the stratified flow in the cold leg and the fluid flow separation and its oscillations in the down comer during a single phase scenario. These physical phenomena are important for the thermal RPV loading assessment. On the other hand, the absence of experimental two-phase data necessitated to carry out an experimental study around the mixing area behavior (free surface, stratified flow) during an ECC injection with an uncovered cold leg. The new EDF R and D mock up, called HYBISCUS, is a facility which is made out of Plexiglas (atmosphere pressure) and represents a half scale CP0 geometry with one cold leg and part of the down comer. The mock up modularity allows us to insert representative ECC nozzles and a thermal shield. In reference to the reactor scenarios, the experimental operating conditions are derived from the conservation of the density effects (Froude number). For that, a heated salted water flow is used to represent the ECC injection whereas water represents the cold leg fluid. This mock up has been defined in order to represent single phase flow (cold leg and down comer full of water) or two-phase flow (uncovered cold leg) ECC scenarios. This paper reports experimental results

  13. Reactor flux calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lhuillier, D. [Commissariat à l' Énergie Atomique et aux Énergies Alternatives, Centre de Saclay, IRFU/SPhN, 91191 Gif-sur-Yvette (France)

    2013-02-15

    The status of the prediction of reactor anti-neutrino spectra is presented. The most accurate method is still the conversion of total β spectra of fissionning isotopes as measured at research reactors. Recent re-evaluations of the conversion process led to an increased predicted flux by few percent and were at the origin of the so-called reactor anomaly. The up to date predictions are presented with their main sources of error. Perspectives are given on the complementary ab-initio predictions and upcoming experimental cross-checks of the predicted spectrum shape.

  14. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  15. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  16. Development of a Tritium Extruder for ITER Pellet Injection

    Energy Technology Data Exchange (ETDEWEB)

    M.J. Gouge; P.W. Fisher

    1998-09-01

    As part of the International Thermonuclear Experimental Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first-wall tritium inventories by a process of "isotopic fueling" in which tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge. This repeating single-stage pneumatic pellet injector, called the Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector, has a piston-driven mechanical extruder and is designed to extrude and accelerate hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and deuterium-tritium (D-T) mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and D-T extrusions; integrate, test, and evaluate the extruder in a repeating, single-stage light gas gun that is sized for the ITER application (pellet diameter -7 to 8 mm); evaluate options for recycling propellant and extruder exhaust gas; and evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory that requires secondary and room containment systems. In tests with deuterium feed at ORNL, up to 13 pellets per extrusion have been extruded at rates up to 1 Hz and accelerated to speeds of 1.0 to 1.1 km/s, using hydrogen propellant gas at a supply pressure of 65 bar. Initially, deuterium pellets 7.5 mm in diameter and 11 mm in length were produced-the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 10% density perturbation to ITER. Subsequently, the extruder nozzle was modified to produce pellets that are almost 7.5-mm right circular

  17. Iterative initial condition reconstruction

    Science.gov (United States)

    Schmittfull, Marcel; Baldauf, Tobias; Zaldarriaga, Matias

    2017-07-01

    Motivated by recent developments in perturbative calculations of the nonlinear evolution of large-scale structure, we present an iterative algorithm to reconstruct the initial conditions in a given volume starting from the dark matter distribution in real space. In our algorithm, objects are first moved back iteratively along estimated potential gradients, with a progressively reduced smoothing scale, until a nearly uniform catalog is obtained. The linear initial density is then estimated as the divergence of the cumulative displacement, with an optional second-order correction. This algorithm should undo nonlinear effects up to one-loop order, including the higher-order infrared resummation piece. We test the method using dark matter simulations in real space. At redshift z =0 , we find that after eight iterations the reconstructed density is more than 95% correlated with the initial density at k ≤0.35 h Mpc-1 . The reconstruction also reduces the power in the difference between reconstructed and initial fields by more than 2 orders of magnitude at k ≤0.2 h Mpc-1 , and it extends the range of scales where the full broadband shape of the power spectrum matches linear theory by a factor of 2-3. As a specific application, we consider measurements of the baryonic acoustic oscillation (BAO) scale that can be improved by reducing the degradation effects of large-scale flows. In our idealized dark matter simulations, the method improves the BAO signal-to-noise ratio by a factor of 2.7 at z =0 and by a factor of 2.5 at z =0.6 , improving standard BAO reconstruction by 70% at z =0 and 30% at z =0.6 , and matching the optimal BAO signal and signal-to-noise ratio of the linear density in the same volume. For BAO, the iterative nature of the reconstruction is the most important aspect.

  18. Conformable variational iteration method

    Directory of Open Access Journals (Sweden)

    Omer Acan

    2017-02-01

    Full Text Available In this study, we introduce the conformable variational iteration method based on new defined fractional derivative called conformable fractional derivative. This new method is applied two fractional order ordinary differential equations. To see how the solutions of this method, linear homogeneous and non-linear non-homogeneous fractional ordinary differential equations are selected. Obtained results are compared the exact solutions and their graphics are plotted to demonstrate efficiency and accuracy of the method.

  19. ITER: Promises unkept ? (1/2)

    CERN Multimedia

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  20. ITER: Promises unkept ? (2/2)

    CERN Multimedia

    CERN. Geneva

    2008-01-01

    Fusion power as the source of energy on Earth has been the dream of mankind ever since the principles were understood. ITER, the Latin word for “the way”, is the world’s largest Fusion device presently under construction in Cadarache, France. Supported by the People’s Republic of China, the European Atomic Energy Community, India, Japan, the Republic of Korea, the Russian Federation, and the United States of America, an international organization was founded after the signature of the Joint ITER Agreement in October of 2006. The goal is to build a Fusion reactor with a power amplification of 10, a total fusion power of 500 MW or more operating at extended burn times of 400-3000 seconds, with Deuterium and Tritium as its basic fuel. Following a short introduction into fusion science principles, the history of thermo nuclear fusion will be covered. Finally more recent construction projects around the world, their latest achievements and the path to ITER will be described. Technological and scientific c...

  1. Iterative robust adaptive beamforming

    Science.gov (United States)

    Li, Yang; Ma, Hong; Cheng, Li

    2017-12-01

    The minimum power distortionless response beamformer has a good interference rejection capability, but the desired signal will be suppressed if signal steering vector or data covariance matrix is not precise. The worst-case performance optimization-based robust adaptive beamformer (WCB) has been developed to solve this problem. However, the solution of WCB cannot be expressed in a closed form, and its performance is affected by a prior parameter, which is the steering vector error norm bound of the desired signal. In this paper, we derive an approximate diagonal loading expression of WCB. This expression reveals a feedback loop relationship between steering vector and weight vector. Then, a novel robust adaptive beamformer is developed based on the iterative implementation of this feedback loop. Theoretical analysis indicates that as the iterative step increases, the performance of the proposed beamformer gets better and the iteration converges. Furthermore, the proposed beamformer does not subject to the steering vector error norm bound constraint. Simulation examples show that the proposed beamformer has better performance than some classical and similar beamformers.

  2. Neutron cameras for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P. [ITER San Diego Joint Work Site, La Jolla, CA (United States)] [and others

    1998-12-31

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from {sup 16}N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with {sup 16}N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins.

  3. Local Neutron Flux Distribution Measurements by Wire-Dosimetry in the AMMON Experimental Program in the EOLE Reactor

    Directory of Open Access Journals (Sweden)

    Gruel A.

    2016-01-01

    Full Text Available Dosimetry measurements were carried out during the AMMON experimental program, in the EOLE facility. Al-0.1 wt% Au wires were positioned along curved fuel plates of JHR-type assemblies to investigate the azimuthal and axial gold capture rate profiles, directly linked to the thermal and epithermal flux. After irradiation, wires were cut into small segments (a few mm, and the gold capture rate of each part was measured by gamma spectrometry on the MADERE platform. This paper presents results in the “hafnium” configuration, and more specifically the azimuthal flux profile characterization. The final uncertainty on each measured wire lies below 1% (at 2 standard deviations. Experimental profiles are in a good agreement against Monte Carlo calculations, and the 4% capture rate increase at the plate edge is well observed. The flux dissymmetry due to assembly position in the core is also measured, and shows a 10% discrepancy between the two edges of the plate.

  4. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  5. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  6. Analysis and interpretation of residence time distribution experimental curves in FM01-LC reactor using axial dispersion and plug dispersion exchange models with closed-closed boundary conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rivera, Fernando F. [Departamento de Quimica, Universidad Autonoma Metropolitana-Iztapalapa, San Rafael Atlixco 186, C.P. 09340, Mexico, D.F. (Mexico); Cruz-Diaz, Martin R., E-mail: mcruz@tese.edu.m [Division de Quimica y Bioquimica, Tecnologico de Estudios Superiores de Ecatepec, Av. Tecnologico S/N Esq. Av. Hank Gonzalez, Valle de Anahuac, C.P. 55120, Ecatepec, Edo. de Mex (Mexico); Rivero, Eligio P. [Departamento de Ingenieria y Tecnologia, Universidad Nacional Autonoma de Mexico, Facultad de Estudios Superiores Cuautitlan, Av. Primero de Mayo, Cuautitlan Izcalli, C.P. 54740, Edo. de Mex (Mexico); Gonzalez, Ignacio [Departamento de Quimica, Universidad Autonoma Metropolitana-Iztapalapa, San Rafael Atlixco 186, C.P. 09340, Mexico, D.F. (Mexico)

    2010-12-15

    The liquid phase mixing flow pattern at low (20 < Re < 120) and intermediate liquid flow rate (120 < Re < 400) was studied by means of residence time distribution (RTD) experimental curve in an up-flow Filter Press electrochemical reactor (FM01-LC) bench scale. For this purpose, a plastic turbulence promoter was used with stainless-steel and platinised titanium structural meshes as electrodes in channel configuration. To visualize and determine the mixing flow pattern in the liquid phase, the stimulus-response technique was employed using dextran blue (D{sub M} = 1.058 x 10{sup -11} m{sup 2} s{sup -1}, 25 {sup o}C, in water) as model tracer. A theoretical analysis and approximation RTD experimental curves with axial dispersion model (ADM) and plug dispersion exchange model (PDE), with 'closed-closed vessel' boundary conditions were used in order to establish a better approximation of the axial dispersion, stagnant zones, channelling and by-pass (preference flow) effects present at low and intermediate Re. RTD curves show that the liquid flow pattern in the FM01-LC deviates considerably from axial dispersion model at low Re, where the FM01-LC exhibits large channelling, stagnant zones, and dead zone. The PDE model represents fairly this deviation from ideal flow (less dead zone).

  7. Experimental study on fluid mixing in a fuel subassembly of a fast reactor. Temperature field around heated pin with cross flow

    Energy Technology Data Exchange (ETDEWEB)

    Miyakoshi, Hiroyuki; Kamide, Hideki; Tanaka, Masaaki; Yamamoto, Kazuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    2002-03-01

    High burnup of the core is one of means to reduce the cost of a fast reactor and fuel cycle system. However, it is not enough to investigate thermohydraulics in the core, in which fuel and wrapper tube are deformed due to irradiation under high burnup condition. In this study, sodium experiment was performed to investigate fluid mixing in a wire-wrapped 37-pin subassembly model, which had local blockage and cross flow around the blockage. Such cross flow is one of elements of thermohydraulics in a deformed subassembly. The experimental results is useful to develop numerical simulation method for the deformed subassembly. Seven pins, each had different relative position to the blockage, were heated individually in the experiments. Temperature field in the subassembly was measured. Influences of the flow rate and heater power were also examined. A horizontal cross flow occurred in upstream region toward the blockage. It was observed that the temperature field was influenced by this cross flow. The measured temperature field showed that there was a bypass flow around the blockage, which flowed toward the center of subassembly. The cross flow due to the bypass flow reached the 3rd row of pins from the blockage. The swirl flow, resulted from the spacer wire, also influenced the temperature field. The obtained experimental data will be used to develop and verify a numerical simulation method for a deformed fuel subassembly. (author)

  8. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD-BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt; Steven P. Antal; Michael Z. Podowski

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  9. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  10. An iterative homogenization technique that preserves assembly core exchanges

    Energy Technology Data Exchange (ETDEWEB)

    Mondot, Ph. [Electricite de France, Recherche et Developement, SINETICS, 92 - Clamart (France); Sanchez, R. [CEA Saclay, Direction des Etudes Nucleaires, Service Etude des Reacteurs et de Modelisations Avancees, 91 - Gilf sur Yvette (France)

    2003-07-01

    A new interactive homogenization procedure for reactor core calculations is proposed that requires iterative transport assembly and diffusion core calculations. At each iteration the transport solution of every assembly type is used to produce homogenized cross sections for the core calculation. The converged solution gives assembly fine multigroup transport fluxes that preserve macro-group assembly exchanges in the core. This homogenization avoids the periodic lattice-leakage model approximation and gives detailed assembly transport fluxes without need of an approximated flux reconstruction. Preliminary results are given for a one-dimensional core model. (authors)

  11. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics; Assemblee generale. Reunion technique: la simulation numerique et experimentale appliquee a la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  12. CONVECTION REACTOR

    Science.gov (United States)

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  13. Experience on divertor fuel retention after two ITER-Like Wall campaigns

    Science.gov (United States)

    Heinola, K.; Widdowson, A.; Likonen, J.; Ahlgren, T.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Guillemaut, C.; Jepu, I.; Krat, S.; Lahtinen, A.; Matthews, G. F.; Mayer, M.; Contributors, JET

    2017-12-01

    The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10–15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

  14. Wide area data replication in an ITER-relevant data environment

    Energy Technology Data Exchange (ETDEWEB)

    Centioli, C. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati, p.b. 65, 00044 Frascati Rome (Italy); Iannone, F. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati, p.b. 65, 00044 Frascati Rome (Italy)]. E-mail: francesco.iannone@frascati.enea.it; Panella, M. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati, p.b. 65, 00044 Frascati Rome (Italy); Vitale, V. [Associazione EURATOM/ENEA sulla Fusione, Centro Ricerche Frascati, p.b. 65, 00044 Frascati Rome (Italy); Bracco, G. [ENEA-INFO Central Computer and Network Service, Lungotevere Thaon di Revel 76, 00196 Rome (Italy); Guadagni, R. [ENEA-INFO Central Computer and Network Service, Lungotevere Thaon di Revel 76, 00196 Rome (Italy); Migliori, S. [ENEA-INFO Central Computer and Network Service, Lungotevere Thaon di Revel 76, 00196 Rome (Italy); Steffe, M. [ENEA-INFO Central Computer and Network Service, Lungotevere Thaon di Revel 76, 00196 Rome (Italy); Eccher, S. [CASPUR Inter-University Computing Consortium, Via dei Tizii 6B, 00185 Rome (Italy); Maslennikov, A. [CASPUR Inter-University Computing Consortium, Via dei Tizii 6B, 00185 Rome (Italy); Mililotti, M. [CASPUR Inter-University Computing Consortium, Via dei Tizii 6B, 00185 Rome (Italy); Molowny, M. [CASPUR Inter-University Computing Consortium, Via dei Tizii 6B, 00185 Rome (Italy); Palumbo, G. [CASPUR Inter-University Computing Consortium, Via dei Tizii 6B, 00185 Rome (Italy); Carboni, M. [GARR Consortium - Viale Palmiro Togliatti 1625, 00155 Rome (Italy)

    2005-11-15

    The next generation of tokamak experiments will require a new way of approaching data sharing issues among fusion organizations. In the fusion community, many researchers at different worldwide sites will analyse data produced by International Thermonuclear Experimental Reactor (ITER), wherever it will be built. In this context, an efficient availability of the data in the sites where the computational resources are located becomes a major architectural issue for the deployment of ITER computational infrastructure. The approach described in this paper goes beyond the usual site-centric model mainly devoted to granting access exclusively to experimental data stored at the device sites. To this aim, we propose a new data replication architecture relying on a wide area network, based on a Master/Slave model and on synchronization techniques producing mirrored data sites. In this architecture, data replication will affect large databases (TB) as well as large UNIX-like file systems, using open source-based software components, namely MySQL, as database management system, and RSYNC and BBFTP for data transfer. A test-bed has been set up to evaluate the performance of the software components underlying the proposed architecture. The test-bed hardware layout deploys a cluster of four Dual-Xeon Supermicro each with a raid array of 1 TB. High performance network line (1 Gbit over 400 km) provides the infrastructure to test the components on a wide area network. The results obtained will be thoroughly discussed.

  15. ITER Blanket First Wall (WBS 1.6{sub 1}A)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kim, H. G.; Kim, J. H. (and others)

    2008-03-15

    International Thermonuclear Experimental Reactor (ITER) project is the international collaboration one for the commercialization of nuclear fusion energy through the technical and engineering verification. In ITER project, we plan to procure the blanket systems which has the risk of technology and cost when it is newly developed. We are developing the manufacturing process and joining technology for the ITER blanket to complete the procurement with qualified blanket system. To evaluate the soundness of manufacturing process, specimen and mock-up tests are being prepared. Finally, we can obtain the key technology of nuclear fusion reactor especially on the blanket design, joining and manufacturing technology through the present project and these technologies will help the construction of Korea fusion DEMO reactor and the development of commercial nuclear fusion reactor in Korea. In 1st year, through the fabrication of the Cu/SS and Be/Cu joint specimen, fabrication procedure such as material preparation, canning, degassing, HIP (Hot Isostatic Pressing), PHHT (Post HIP heat treatment) was established. The optimized HIP conditions (1050 .deg. C, 150 MPa, 2 hr for Cu/SS and 580 - 620 .deg. C, 100-150 MPa, 2 hr for Be/Cu) were developed through the investigation on joint specimen fabricated with the various HIP conditions; the destructive tests of joint and NDT such as UT (10 MHz, 0.25 inch D, flat type) and ECT. Several mock-ups were fabricated for confirming the joint integrity and NDT. specimens fabricated with these mock-ups were used in mechanical tests including microstructure observation. The mock-ups were used in the HHF test after the developed NDT. In 2nd year, PHHT of Cu was investigated in order to recover its mechanical properties, and the pre-qualification mock-up were fabricated against the Qualification Program and sent to RF for HHF testing in TSEFEY. FW fabrication and joining procedure were documented in the form of the TSD. Qualification mock

  16. Experimental Measurements of Heat Transfer through a Lunar Regolith Simulant in a Vibro-Fluidized Reactor Oven

    Science.gov (United States)

    Nayagam, Vedha; Berger, Gordon M.; Sacksteder, Kurt R.; Paz, Aaron

    2012-01-01

    Extraction of mission consumable resources such as water and oxygen from the planetary environment provides valuable reduction in launch-mass and potentially extends the mission duration. Processing of lunar regolith for resource extraction necessarily involves heating and chemical reaction of solid material with processing gases. Vibrofluidization is known to produce effective mixing and control of flow within granular media. In this study we present experimental results for vibrofluidized heat transfer in lunar regolith simulants (JSC-1 and JSC-1A) heated up to 900 C. The results show that the simulant bed height has a significant influence on the vibration induced flow field and heat transfer rates. A taller bed height leads to a two-cell circulation pattern whereas a single-cell circulation was observed for a shorter height. Lessons learned from these test results should provide insight into efficient design of future robotic missions involving In-Situ Resource Utilization.

  17. PREFACE: Progress in the ITER Physics Basis

    Science.gov (United States)

    Ikeda, K.

    2007-06-01

    I would firstly like to congratulate all who have contributed to the preparation of the `Progress in the ITER Physics Basis' (PIPB) on its publication and express my deep appreciation of the hard work and commitment of the many scientists involved. With the signing of the ITER Joint Implementing Agreement in November 2006, the ITER Members have now established the framework for construction of the project, and the ITER Organization has begun work at Cadarache. The review of recent progress in the physics basis for burning plasma experiments encompassed by the PIPB will be a valuable resource for the project and, in particular, for the current Design Review. The ITER design has been derived from a physics basis developed through experimental, modelling and theoretical work on the properties of tokamak plasmas and, in particular, on studies of burning plasma physics. The `ITER Physics Basis' (IPB), published in 1999, has been the reference for the projection methodologies for the design of ITER, but the IPB also highlighted several key issues which needed to be resolved to provide a robust basis for ITER operation. In the intervening period scientists of the ITER Participant Teams have addressed these issues intensively. The International Tokamak Physics Activity (ITPA) has provided an excellent forum for scientists involved in these studies, focusing their work on the high priority physics issues for ITER. Significant progress has been made in many of the issues identified in the IPB and this progress is discussed in depth in the PIPB. In this respect, the publication of the PIPB symbolizes the strong interest and enthusiasm of the plasma physics community for the success of the ITER project, which we all recognize as one of the great scientific challenges of the 21st century. I wish to emphasize my appreciation of the work of the ITPA Coordinating Committee members, who are listed below. Their support and encouragement for the preparation of the PIPB were

  18. Runaway electrons and ITER

    Science.gov (United States)

    Boozer, Allen H.

    2017-05-01

    The potential for damage, the magnitude of the extrapolation, and the importance of the atypical—incidents that occur once in a thousand shots—make theory and simulation essential for ensuring that relativistic runaway electrons will not prevent ITER from achieving its mission. Most of the theoretical literature on electron runaway assumes magnetic surfaces exist. ITER planning for the avoidance of halo and runaway currents is focused on massive-gas or shattered-pellet injection of impurities. In simulations of experiments, such injections lead to a rapid large-scale magnetic-surface breakup. Surface breakup, which is a magnetic reconnection, can occur on a quasi-ideal Alfvénic time scale when the resistance is sufficiently small. Nevertheless, the removal of the bulk of the poloidal flux, as in halo-current mitigation, is on a resistive time scale. The acceleration of electrons to relativistic energies requires the confinement of some tubes of magnetic flux within the plasma and a resistive time scale. The interpretation of experiments on existing tokamaks and their extrapolation to ITER should carefully distinguish confined versus unconfined magnetic field lines and quasi-ideal versus resistive evolution. The separation of quasi-ideal from resistive evolution is extremely challenging numerically, but is greatly simplified by constraints of Maxwell’s equations, and in particular those associated with magnetic helicity. The physics of electron runaway along confined magnetic field lines is clarified by relations among the poloidal flux change required for an e-fold in the number of electrons, the energy distribution of the relativistic electrons, and the number of relativistic electron strikes that can be expected in a single disruption event.

  19. Helium processing for deuterium/helium burns in ITER's physics phase

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.; Sze, D.K.

    1991-01-01

    The requirements for vacuum pumping and fuel processing for deuterium/helium (D/{sup 3}He) burns in the physics operating phase for the International Thermonuclear Experimental Reactor (ITER) were assessed. These burns are expected to have low fusion power (100 MW), short burn times ({le}30 s), limited operation (2000 shots), and a fractional burn {approximately}0.3%. For the physics phase, the fuel processing system will include several units to separate deuterium and helium (activated charcoal bed, SAES getter and a Pd/Ag diffuser), as well as an isotopic separation system to separate {sup 3}He and {sup 4}He. The needed vacuum system's cryosorption surface area may be as large as 10 m{sup 2} if the burn time is {approximately}200 s, the fractional burn is <0.3%, or the fusion power is >100 MW. 8 refs., 1 fig., 4 tabs.

  20. Searching with iterated maps

    Science.gov (United States)

    Elser, V.; Rankenburg, I.; Thibault, P.

    2007-01-01

    In many problems that require extensive searching, the solution can be described as satisfying two competing constraints, where satisfying each independently does not pose a challenge. As an alternative to tree-based and stochastic searching, for these problems we propose using an iterated map built from the projections to the two constraint sets. Algorithms of this kind have been the method of choice in a large variety of signal-processing applications; we show here that the scope of these algorithms is surprisingly broad, with applications as diverse as protein folding and Sudoku. PMID:17202267

  1. Initial design and test of the tritium breeder monitoring system for the lead-lithium cooled ceramic breeder (LLCB) module of the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kapyshev, V., E-mail: kapyshev@nikiet.ru [Federal State Unitary Enterprise “Dollezhal Research and Development Institute of Power Engineering”, PO Box 788, Moscow 101000 (Russian Federation); Danilov, I.; Kartashev, I.; Kovalenko, V.; Leshukov, A.; Poliksha, V.; Razmerov, A.; Strebkov, Yu.; Sviridenko, M.; Trusova, E.; Vladimirova, N. [Federal State Unitary Enterprise “Dollezhal Research and Development Institute of Power Engineering”, PO Box 788, Moscow 101000 (Russian Federation); Kalashnikov, A. [Rosatom (Russian Federation)

    2013-10-15

    Highlights: • We propose a system for measurement of tritium-breeding dynamics in module of the ITER. • Lithium carbonate sensors and neutron detectors are used for the measurements. • The sensors and detectors are irradiated by neutrons under ITER plasma operations. • A pneumatic concept is suggested for conveying of the samples in the module. • The results of the tritium measurements in the sensors after irradiation are discussed. -- Abstract: The demonstration of a tritium breeder is an important part of an ITER mission. A concept for an experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a test breeder module (TBM) of ITER has been developed. A system for the experimental estimation of the values is proposed for the lead–lithium cooled ceramic breeder (LLCB) TBM of ITER. The system is based on tritium breeder and neutron flux measurements under ITER plasma D-T experiments and the use of lithium carbonate sensors and neutron detectors. Three capsules with lithium carbonate (Li{sub 2}CO{sub 3}) containing different relative abundances of lithum-6 and lithium-7 (Li-6/Li-7 ratios of that found in nature: 1/1 and 9/1) and three capsules with neutron detectors are placed in a container. The low-level activities of the structural materials of the container and the capsules are used to prevent operation in a hot cell after the reactor irradiation of the container. For the delivery/withdrawal of the containers into/from the TBM, a pneumatic concept is suggested with a monitor channel connecting the TBM and an operating zone for conveying the containers in the TBM before the pulse and extracting them after the pulse. The initial design of the container with the capsules for the samples and the channel part in the TBM are presented in this paper. A laboratory facility for the investigation of the pneumatic parameters and the container moving in the channel is proposed. Neutron calculation is performed to estimate

  2. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: Application to the treatment of experimental oral cancer

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, E. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)], E-mail: epozzi@cnea.gov.ar; Nigg, D.W. [Idaho National Laboratory, Idaho Falls (United States); Miller, M.; Thorp, S.I. [Instrumentation and Control Department, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Heber, E.M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Zarza, L.; Estryk, G. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Monti Hughes, A.; Molinari, A.J.; Garabalino, M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Itoiz, M.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Aromando, R.F. [Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Quintana, J. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Trivillin, V.A.; Schwint, A.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)

    2009-07-15

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1x10{sup 9} n cm{sup -2} s{sup -1} and the fast neutron flux was 2.5x10{sup 6} n cm{sup -2} s{sup -1}, indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in {sup 6}Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  3. Combined coagulation-flocculation and sequencing batch reactor with phosphorus adjustment for the treatment of high-strength landfill leachate: experimental kinetics and chemical oxygen demand fractionation.

    Science.gov (United States)

    El-Fadel, M; Matar, F; Hashisho, J

    2013-05-01

    The treatability of high-strength landfill leachate is challenging and relatively limited. This study examines the feasibility of treating high-strength landfill leachate (chemical oxygen demand [COD]: 7,760-11,770 mg/L, biochemical oxygen demand [BOD5]: 2,760-3,569 mg/L, total nitrogen [TN] = 980-1,160 mg/L) using a sequencing batch reactor (SBR) preceded by a coagulation-flocculation process with phosphorus nutritional balance under various mixing and aeration patterns. Simulations were also conducted to define kinetic parameters and COD fractionation. Removal efficiencies reached 89% for BOD5, 60% for COD, and 72% for TN, similar to and better than reported studies, albeit with a relatively lower hydraulic retention time (HRT) and solid retention time (SRT). The coupled experimental and simulation results contribute in filling a gap toward managing high-strength landfill leachate and providing guidelines for corresponding SBR applications. The treatability of high-strength landfill leachate, which is challenging and relatively limited, was demonstrated using a combined coagulation-flocculation with SBR technology and nutrient balance adjustment. The most suitable coagulant, kinetic design parameters, and COD fractionation were defined using coupled experimental and simulation results contributing in filling a gap toward managing high-strength leachate by providing guidelines for corresponding SBR applications and anticipating potential constraints related to the non-biodegradable COD fraction. In this context, while the combined coagulation-flocculation and SBR process improved removal efficiencies, posttreatment may be required for high-strength leachate, depending on discharge standards and ultimate usage of the treated leachate.

  4. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: application to the treatment of experimental oral cancer.

    Science.gov (United States)

    Pozzi, E; Nigg, D W; Miller, M; Thorp, S I; Heber, E M; Zarza, L; Estryk, G; Monti Hughes, A; Molinari, A J; Garabalino, M; Itoiz, M E; Aromando, R F; Quintana, J; Trivillin, V A; Schwint, A E

    2009-07-01

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1 x 10(9) n cm(-2)s(-1) and the fast neutron flux was 2.5 x 10(6) n cm(-2)s(-1), indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in (6)Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  5. Design of a DC Busbar for the ITER PF Converter

    Science.gov (United States)

    Guo, Bin; Song, Zhiquan; Xu, Liuwei; Zhang, Ming; Li, Jinchao; Jiang, Li; Fu, Peng; Wang, Min; Dong, Lin

    2014-04-01

    The DC busbar is an important component for the ITER PF converter module to connect the converter and the reactor. This paper analyzes different cross-sections and different thermodynamic properties under natural-cooling and water-cooling conditions, and simulation is carried out by the software of the finite element method (FEM). The result of the analysis shows that the water-cooling method is the better choice for the DC busbar.

  6. Influence of the optical fiber type on the performances of fiber-optics current sensor dedicated to plasma current measurement in ITER.

    Science.gov (United States)

    Aerssens, Matthieu; Descamps, Frédéric; Gusarov, Andrei; Mégret, Patrice; Moreau, Philippe; Wuilpart, Marc

    2015-07-01

    In this paper, we compare, by means of simulations using the Jones formalism, the performances of several optical fiber types (low birefringence and spun fibers) for the measurement of plasma current in international thermonuclear experimental reactor (ITER). The main results presented in this paper concern the minimum value of the ratio between the beat length and the spun period, which allows meeting the ITER current measurement specifications. Assuming a high-birefringence spun fiber with a beat length of 3 mm, we demonstrate that the minimum ratio between the beat length and the spun period is 4.4 when considering a 28 m long sensing fiber surrounding the vacuum vessel. This minimum ratio rises to 10.14 when a 100 m long lead fiber connecting the interrogating system to the sensing fiber is taken into account.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

  8. Efficient simulation of periodically forced reactors with radial gradients

    NARCIS (Netherlands)

    van de Rotten, Bart A.; Verduyn Lunel, Sjoerd M.; Bliek, A.

    2006-01-01

    The aim of this paper is to present a limited memory iterative method, called the Broyden Rank Reduction method, to simulate periodically forced processes in plug-flow reactors with radial gradients taken into account. The simulation of periodically forced processes in plug-flow reactors leads to

  9. 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

  10. REACTOR COOLING

    Science.gov (United States)

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  11. Iterative cross section sequence graph for handwritten character segmentation.

    Science.gov (United States)

    Dawoud, Amer

    2007-08-01

    The iterative cross section sequence graph (ICSSG) is an algorithm for handwritten character segmentation. It expands the cross section sequence graph concept by applying it iteratively at equally spaced thresholds. The iterative thresholding reduces the effect of information loss associated with image binarization. ICSSG preserves the characters' skeletal structure by preventing the interference of pixels that causes flooding of adjacent characters' segments. Improving the structural quality of the characters' skeleton facilitates better feature extraction and classification, which improves the overall performance of optical character recognition (OCR). Experimental results showed significant improvements in OCR recognition rates compared to other well-established segmentation algorithms.

  12. Nuclear heating experiments for the validation of fusion reactor shielding performance

    Energy Technology Data Exchange (ETDEWEB)

    Batistoni, P. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Angelone, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Pillon, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Rado, V. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy)

    1997-05-01

    Nuclear heating experiments were performed at the 14 MeV Frascati neutron generator, with the purpose of validating the shielding performance of the International Thermonuclear Experiment Reactor (ITER) shielding system (first wall, shielding blanket and vacuum vessel). The experiments consisted of the irradiation, with 14 MeV neutrons, of large assemblies that simulated, for size and material composition, the shielding system and the toroidal field magnet. The nuclear heating was measured with different thermo-luminescent detectors, as a function of the penetration depth inside the assemblies. The experimental results obtained were compared with the same quantities calculated with the same nuclear database and code used for nuclear heating calculations in the ITER design. (orig.)

  13. Radioactivity measurements of ITER materials using the TFTR D-T neutron field

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A.; Abdou, M.A. [California Univ., Los Angeles, CA (United States). School of Engineering and Applied Science; Barnes, C.W. [Los Alamos National Lab., NM (United States); Kugel, H.W. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Loughlin, M.J. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

    1994-08-01

    The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials. for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc. zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.

  14. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

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

  15. The probability of iterated conditionals

    NARCIS (Netherlands)

    van Wijnbergen-Huitink, Janneke; Elqayam, Shira; Over, David E.

    Iterated conditionals of the form If p, then if q, r are an important topic in philosophical logic. In recent years, psychologists have gained much knowledge about how people understand simple conditionals, but there are virtually no published psychological studies of iterated conditionals. This

  16. Thermal Dissipation Modelling and Design of ITER PF Converter Alternating Current Busbar

    Science.gov (United States)

    Guo, Bin; Song, Zhiquan; Fu, Peng; Jiang, Li; Li, Jinchao; Wang, Min; Dong, Lin

    2016-10-01

    Because the larger metallic surrounds are heated by the eddy current, which is generated by the AC current flowing through the AC busbar in the International Thermonuclear Experimental Reactor (ITER) poloidal field (PF) converter system, shielding of the AC busbar is required to decrease the temperature rise of the surrounds to satisfy the design requirement. Three special types of AC busbar with natural cooling, air cooling and water cooling busbar structure have been proposed and investigated in this paper. For each cooling scheme, a 3D finite model based on the proposed structure has been developed to perform the electromagnetic and thermal analysis to predict their operation behavior. Comparing the analysis results of the three different cooling patterns, water cooling has more advantages than the other patterns and it is selected to be the thermal dissipation pattern for the AC busbar of ITER PF converter unit. The approach to qualify the suitable cooling scheme in this paper can be provided as a reference on the thermal dissipation design of AC busbar in the converter system. supported by National Natural Science Foundation of China (No. 51407179)

  17. Optimisation of hot isostatic pressing bonded SS/SS joints conditions for ITER blanket shield

    Energy Technology Data Exchange (ETDEWEB)

    Cedat, D. [AREVA NP Technical Centre 30, Bld de l' industrie, 71205 Le Creusot Cedex (France); Bobin, I., E-mail: isabelle.bobinvastra@areva.com [AREVA NP Technical Centre 30, Bld de l' industrie, 71205 Le Creusot Cedex (France); Boireau, B. [AREVA NP Technical Centre 30, Bld de l' industrie, 71205 Le Creusot Cedex (France); Bucci, P. [CEA DRT/Liten/DTH, 38054 Grenoble (France); Lorenzetto, P. [FUSION FOR ENERGY, Torres Diagonal Litoral, B3, Carrer Josep Pla, 2, 08019 Barcelona (Spain)

    2012-08-15

    In the engineering design activity of international thermonuclear experimental reactor (ITER), stainless steels are being considered as candidates materials for several module type structures. Hot isostatic pressing (HIP) technique is expected for the fabrication of these modules. Stainless steel powders are simultaneously consolidated as mono-material block or/and joined in bi-material module. This paper reviews the manufacturing stages, non-destructive examination and the developments of the HIP bonded joints of 316L SS (powder and solid) for application to the ITER shield blanket. It is well known that the powder surface oxidation negatively influences the impact toughness of raw material and joints consolidated by this way. In order to get acceptable mechanical properties of materials, a study on the effect of reducing the powder oxygen content has been launched. To evaluate susceptibility to the oxygen content of HIPed joint specimens, tensile and toughness tests have been performed. From this study, optimal conditions of HIP were fitted and the influence of oxygen was mastered to obtain good mechanical properties of the consolidated powder material as well as for HIPed junction.

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

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

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

  19. ITER perspective on fusion reactor diagnostics - A spectroscopic view

    DEFF Research Database (Denmark)

    De Bock, M. F. M.; Barnsley, R.; Bassan, M.

    2016-01-01

    challenges to the development of spectroscopic (but also other) diagnostics. This contribution presents an overview of recent achievements in 4 topical areas: First mirror protection and cleaning, Nuclear confinement, Radiation mitigation strategy for optical and electronic components and Calibration...

  20. Advances in iterative multigrid PIV image processing

    Science.gov (United States)

    Scarano, F.; Riethmuller, M. L.

    2000-12-01

    An image-processing technique is proposed, which performs iterative interrogation of particle image velocimetry (PIV) recordings. The method is based on cross-correlation, enhancing the matching performances by means of a relative transformation between the interrogation areas. On the basis of an iterative prediction of the tracers motion, window offset and deformation are applied, accounting for the local deformation of the fluid continuum. In addition, progressive grid refinement is applied in order to maximise the spatial resolution. The performances of the method are analysed and compared with the conventional cross correlation with and without the effect of a window discrete offset. The assessment of performance through synthetic PIV images shows that a remarkable improvement can be obtained in terms of precision and dynamic range. Moreover, peak-locking effects do not affect the method in practice. The velocity gradient range accessed with the application of a relative window deformation (linear approximation) is significantly enlarged, as confirmed in the experimental results.

  1. Logistics management for storing multiple cask plug and remote handling systems in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ventura, Rodrigo, E-mail: rodrigo.ventura@isr.ist.utl.pt [Laboratório de Robótica e Sistemas em Engenharia e Ciência – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Ferreira, João, E-mail: jftferreira@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal); Filip, Iulian, E-mail: ifilip@gmail.com [Faculty of Mechanical Engineering – Technical University Gheorghe Asachi of Iasi, 61 Dimitrie Mangeron Bldv., Iasi 700050 (Romania); Vale, Alberto, E-mail: avale@ipfn.ist.utl.pt [Instituto de Plasmas e Fusão Nuclear – Laboratório Associado, Instituto Superior Técnico, Universidade Técnica de Lisboa, Av. Rovisco Pais 1, 1049-001 Lisboa (Portugal)

    2013-10-15

    Highlights: ► We model the logistics management problem in ITER, taking into account casks of multiple typologies. ► We propose a method to determine the best position of the casks inside a given storage area. ► Our method obtains the sequence of operations required to retrieve or store an arbitrary cask, given its storage place. ► We illustrate our method with simulation results in an example scenario. -- Abstract: During operation, maintenance inside the reactor building at ITER (International Thermonuclear Experimental Reactor) has to be performed by remote handling, due to the presence of activated materials. Maintenance operations involve the transportation and storage of large, heavyweight casks from and to the tokamak building. The transportation is carried out by autonomous vehicles that lift and move beneath these casks. The storage of these casks face several challenges, since (1) the cask storage area is limited in space, and (2) all casks have to be accessible for transportation by the vehicles. In particular, casks in the storage area may block other casks, so that the former has to be moved to a temporary position to give way to the latter. This paper addresses the challenge of managing the logistics of cask storage, where casks may have different typologies. In particular, we propose an approach to (1) determine the best position of the casks inside the storage area, and to (2) obtain the sequence of operations required to retrieve and store an arbitrary cask from/to a given storage place. A combinatorial optimization approach is used to obtain solutions to both these problems. Simulation results illustrate the application of the proposed method to a simple scenario.

  2. On monitoring the tritium breeder in a lead-lithium cooled ceramic breeder (LLCB) module of the ITER

    Energy Technology Data Exchange (ETDEWEB)

    Kapyshev, V., E-mail: kapyshev@nikiet.ru [Federal State Unitary Enterprise ' Dollezhal Research and Development Institute of Power Engineering' , PO Box 788, Moscow 101000 (Russian Federation); Kartashev, I.; Kovalenko, V.; Leshukov, A.; Poliksha, V.; Rasmerov, A.; Strebkov, Yu.; Yukhnov, N.; Vladimirova, N. [Federal State Unitary Enterprise ' Dollezhal Research and Development Institute of Power Engineering' , PO Box 788, Moscow 101000 (Russian Federation)

    2012-08-15

    The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead-lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D-T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method. Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.

  3. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.; Reiter, D. [Forschungszentrum Juelich (DE). Inst. fuer Energieforschung (IEF), Plasmaphysik (IEF-4); Kukushkin, A.S. [ITER International Team, Cadarache (France)

    2007-11-15

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - {approx} 10{sup 21} m{sup -3}, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non

  4. Gap study on technical standards and quality assurance between ITER and Korean regulation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myoung-suk [Kyung Hee University, Yongin-si 446-701, Gyeonggi-do (Korea, Republic of); Lee, Sangil; Lee, Hyoen Gon [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Yongin-si 446-701, Gyeonggi-do (Korea, Republic of)

    2016-11-01

    Highlights: • This paper is to secure knowledge and expand domestic specialists for license and regulation of future fusion facilities including K-DEMO. - Abstract: Currently, Korea Domestic Agency (KODA) for the International Thermonuclear Experimental Reactor (ITER) project or the ITER Korea has been conducting follow-up surveys and development researches for the core technologies of its non-procurement items. This eventually aims to secure implicit as well as explicit knowledge and expand domestic specialists for future fusion facilities including K-DEMO. As a part of the project, the comparative analysis and gap study of technical standards and Quality Assurance (QA) system between ITER and Korean nuclear facilities are being performed. The goal of this study is to present a roadmap re-systematizing current safety-related standards for the fission-based facilities into fusion-based facilities in Republic of Korea. The research is considering two different tracks. First one is to secure the state-of-the-art of the fusion technical standards continually updated. For its investigation, we organized the Korean fusion safety advisory group that consists of experts from institutes, industries, and universities. The latest information of technical standards have been gathered and updated from each area of expert in periodic manner. The information is classified into Product Breakdown Structure (PBS) including safety classification, major safety issues, analysis computer codes, etc. This task will be re-organized this year such that the technical standards for procurement as well as non-procurement items can be accumulated in systematic and sustainable manner. Second one, called “Gap Study,” is to compare the requirements for domestic nuclear facilities and the results from the first track on the basis of experts’ opinion from the national and private sectors and to recommend the approaches to overcome technical and administrative barriers. The result is

  5. Analysis of the Hydrogen Reduction Rate of Magnetite Concentrate Particles in a Drop Tube Reactor Through CFD Modeling

    Science.gov (United States)

    Fan, Deqiu; Mohassab, Yousef; Elzohiery, Mohamed; Sohn, H. Y.

    2016-06-01

    A computational fluid dynamics (CFD) approach, coupled with experimental results, was developed to accurately evaluate the kinetic parameters of iron oxide particle reduction. Hydrogen reduction of magnetite concentrate particles was used as a sample case. A detailed evaluation of the particle residence time and temperature profile inside the reactor is presented. This approach eliminates the errors associated with assumptions like constant particle temperature and velocity while the particles travel down a drop tube reactor. The gas phase was treated as a continuum in the Eulerian frame of reference, and the particles are tracked using a Lagrangian approach in which the trajectory and velocity are determined by integrating the equation of particle motion. In addition, a heat balance on the particle that relates the particle temperature to convection and radiation was also applied. An iterative algorithm that numerically solves the governing coupled ordinary differential equations was developed to determine the pre-exponential factor and activation energy that best fit the experimental data.

  6. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  7. The European contribution to the ITER Remote Maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Damiani, C., E-mail: carlo.damiani@f4e.europa.eu [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Annino, C.; Balagué, S.; Bates, P.; Ceccanti, F.; Di Mascio, T.; Dubus, G.; Esqué, S.; Gonzalez, C.; Lewczanin, M.; Locke, D.; Mont, L.; Olajos, K.; Ranz, R.; Shuff, R.; Puiu, A.; Van Hille, C.; Van Uffelen, M. [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Choi, C.H.; Friconneau, J.P. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2014-10-15

    Highlights: •The article introduces the needs for remote maintenance in ITER. •It also discusses some of the issues related to the cultural transition from tokamaks as plasma physics to nuclear reactors. •It highlights the related cultural change and the implications on plant topology and maintenance. •Then, it presents those remote handling systems that will be procured by Europe. •The article emphasises the need of a major involvement of industries from now on. -- Abstract: For a first-of-a-kind nuclear fusion reactor like ITER, remote maintainability of neutron-activated components is one of the key aspects of plant design and operations, and a fundamental ingredient for the demonstration of long-term viability of fusion as energy source. The European Domestic Agency (EU DA, i.e. Fusion for Energy, F4E) is providing important support to the ITER Organisation (IO) in specifying the functional requirements of the Remote Handling (RH) Procurement Packages (i.e. the subsystems allocated to EU DA belonging to the overall ITER Remote Maintenance Systems IRMS), and in performing design and R and D activities – with the support of national laboratories and industries – in order to define a sound concept for these packages. Furthermore, domestic industries are being involved in the subsequent detailed design, validation, manufacturing and installation activities, in order to actually fulfil our procurement-in-kind obligations. After an introduction to ITER Remote Maintenance, this paper will present status and next stages for the RH systems allocated to EU DA, and will also illustrate complementary aspects related to cross cutting technologies like radiation tolerant components and RH control systems. Finally, the way all these efforts are coordinated will be presented together with the overall implementation scenario and key milestones.

  8. Iterative optimization in inverse problems

    CERN Document Server

    Byrne, Charles L

    2014-01-01

    Iterative Optimization in Inverse Problems brings together a number of important iterative algorithms for medical imaging, optimization, and statistical estimation. It incorporates recent work that has not appeared in other books and draws on the author's considerable research in the field, including his recently developed class of SUMMA algorithms. Related to sequential unconstrained minimization methods, the SUMMA class includes a wide range of iterative algorithms well known to researchers in various areas, such as statistics and image processing. Organizing the topics from general to more

  9. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  10. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  11. NUCLEAR REACTOR

    Science.gov (United States)

    Grebe, J.J.

    1959-07-14

    High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

  12. Rollout sampling approximate policy iteration

    NARCIS (Netherlands)

    Dimitrakakis, C.; Lagoudakis, M.G.

    2008-01-01

    Several researchers have recently investigated the connection between reinforcement learning and classification. We are motivated by proposals of approximate policy iteration schemes without value functions, which focus on policy representation using classifiers and address policy learning as a

  13. Cooperation between CERN and ITER

    CERN Multimedia

    2008-01-01

    CERN and the International Fusion Organisation ITER have just signed a first cooperation agreeement. Kaname Ikeda, the Director-General of the International Fusion Energy Organisation (ITER) (on the right) and Robert Aymar, Director-General of CERN, signing the agreement.The Director-General of the International Fusion Energy Organization, Mr Kaname Ikeda, and CERN Director-General, Robert Aymar, signed a cooperation agreement at a meeting on the Meyrin site on Thursday 6 March. One of the main purposes of this agreement is for CERN to give ITER the benefit of its experience in the field of technology as well as in administrative domains such as finance, procurement, human resources and informatics through the provision of consultancy services. Currently in its start-up phase at its Cadarache site, 70 km from Marseilles (France), ITER will focus its research on the scientific and technical feasibility of using fusion energy as a fu...

  14. Development of liquid metal type TBM technology for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; Kwak, J. G.; Kim, Y. (and others)

    2008-03-15

    The objectives of the ITER project for the construction and operation are to perform the test related to the neutronics, blanket module, tritium treatment technology, advanced plasma technology, and to test the heat extraction and tritium breeding in the test blanket for the fusion reactor. Other parties have been developing the Test Blanket Module (TBM) for testing in the ITER for these purposes. Through this project, we can secure the TBM design and related technology, which will be used as the core technology for the DEMO construction, our own fusion reactor development. In 1st year, the optimized design procedure was established with the existing tools, which have been used in nuclear reactor design, and the optimized HCML TBM design was obtained through iteration method according to the developed design procedure. He cooling system as a TBM auxiliary system was designed considering the final design of the KO HCML TBM such as coolant capacity and operation pressure. Layout for this system was prepared to be installed in the ITER TCWS vault. MHD effect of liquid Li breeder by magnetic flux in ITER such as much higher pressure drop was evaluated with CFD-ACE and it was concluded that the Li breeder should have a slow velocity to reduce this effect. Most results were arranged in the form of DDD including preliminary safety analysis report. In 2nd year, the optimized design procedure was complemented and updated. In performance analysis on thermal-hydraulic and thermo-mechanical one, full 3D meshes were generated and used in this analysis in order to obtain the more exact temperature, deformation, and stress solution. For liquid Li breeder system, design parameters were induced before the detailed design of the system and were used in the design of the liquid Li test loop. LOCA analysis, activation analysis in LOCA, EM analysis were performed as a preliminary safety analysis. In order to develop the manufacturing technology, Be+FMS and FMS to FMS joining conditions

  15. ITER leader to head CERN

    CERN Multimedia

    Feder, Toni

    2003-01-01

    After successfully chairing an external review committee for CERN last year, Robert Aymar will leave ITER to become director general of the European particle physics laboratory rom 2004. Before ITER he also successfully managed the startup or Tore Supra. He will attempt to ensure that the LHC begins operating in 2007 - two years late - and is paid for by 2010 and will also start the planning for life after the LHC (1 page)

  16. Iterative regularization with minimum-residual methods

    DEFF Research Database (Denmark)

    Jensen, Toke Koldborg; Hansen, Per Christian

    2006-01-01

    subspaces. We provide a combination of theory and numerical examples, and our analysis confirms the experience that MINRES and MR-II can work as general regularization methods. We also demonstrate theoretically and experimentally that the same is not true, in general, for GMRES and RRGMRES - their success......We study the regularization properties of iterative minimum-residual methods applied to discrete ill-posed problems. In these methods, the projection onto the underlying Krylov subspace acts as a regularizer, and the emphasis of this work is on the role played by the basis vectors of these Krylov...... as regularization methods is highly problem dependent....

  17. Iterative Regularization with Minimum-Residual Methods

    DEFF Research Database (Denmark)

    Jensen, Toke Koldborg; Hansen, Per Christian

    2007-01-01

    subspaces. We provide a combination of theory and numerical examples, and our analysis confirms the experience that MINRES and MR-II can work as general regularization methods. We also demonstrate theoretically and experimentally that the same is not true, in general, for GMRES and RRGMRES their success......We study the regularization properties of iterative minimum-residual methods applied to discrete ill-posed problems. In these methods, the projection onto the underlying Krylov subspace acts as a regularizer, and the emphasis of this work is on the role played by the basis vectors of these Krylov...... as regularization methods is highly problem dependent....

  18. Experimental data and numerical predictions of a single-phase flow in a batch square stirred tank reactor with a rotating cylinder agitator

    Science.gov (United States)

    Escamilla-Ruíz, I. A.; Sierra-Espinosa, F. Z.; García, J. C.; Valera-Medina, A.; Carrillo, F.

    2017-09-01

    Single-phase flows in stirred tank reactors have useful characteristics for a wide number of industrial applications. Usually, reactors are cylindrical vessels and complex impeller designs, which are often highly energy consuming and produce complicated flow patterns. Therefore, a novel configuration consisting of a square stirred tank reactor is proposed in this study with potential advantages over conventional reactors. In the present work hydrodynamics and turbulence have been studied for a single-phase flow in steady state operating in batch condition. The flow was induced by drag from a rotating cylinder with two diameters. The effects of drag from the stirrer as well as geometrical parameters of the system on the hydrodynamic behavior were investigated using Computational Fluids Dynamics (CFD) and non-intrusive Laser Doppler Anemometry, (LDA). Data obtained from LDA measurements were used for the validation of the CFD simulations, and to detecting the macro-instabilities inside the tank, based on the time series analysis for three rotational speeds N = 180, 1000 and 2000 rpm. The numerical results revealed the formation of flow patterns and macro-vortex structures in the upper part of the tank as consequence of the Reynolds number and the stream discharge emanated from the cylindrical stirrer. Moreover, increasing the cylinder diameter has an impact on the number of recirculation loops as well as the energy consumption of the entire system showing better performance in the presence of turbulent flows.

  19. ITER Central Solenoid Module Fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Smith, John [General Atomics, San Diego, CA (United States)

    2016-09-23

    The fabrication of the modules for the ITER Central Solenoid (CS) has started in a dedicated production facility located in Poway, California, USA. The necessary tools have been designed, built, installed, and tested in the facility to enable the start of production. The current schedule has first module fabrication completed in 2017, followed by testing and subsequent shipment to ITER. The Central Solenoid is a key component of the ITER tokamak providing the inductive voltage to initiate and sustain the plasma current and to position and shape the plasma. The design of the CS has been a collaborative effort between the US ITER Project Office (US ITER), the international ITER Organization (IO) and General Atomics (GA). GA’s responsibility includes: completing the fabrication design, developing and qualifying the fabrication processes and tools, and then completing the fabrication of the seven 110 tonne CS modules. The modules will be shipped separately to the ITER site, and then stacked and aligned in the Assembly Hall prior to insertion in the core of the ITER tokamak. A dedicated facility in Poway, California, USA has been established by GA to complete the fabrication of the seven modules. Infrastructure improvements included thick reinforced concrete floors, a diesel generator for backup power, along with, cranes for moving the tooling within the facility. The fabrication process for a single module requires approximately 22 months followed by five months of testing, which includes preliminary electrical testing followed by high current (48.5 kA) tests at 4.7K. The production of the seven modules is completed in a parallel fashion through ten process stations. The process stations have been designed and built with most stations having completed testing and qualification for carrying out the required fabrication processes. The final qualification step for each process station is achieved by the successful production of a prototype coil. Fabrication of the first

  20. A novel ultra-thin 3D detector—For plasma diagnostics at JET and ITER tokamaks

    Science.gov (United States)

    García, Francisco; Pelligrini, G.; Balbuena, J.; Lozano, M.; Orava, R.; Ullan, M.

    2009-08-01

    A novel ultra-thin silicon detector called U3DTHIN has been designed and built for applications that range from Neutral Particle Analyzers (NPA) used in Corpuscular Diagnostics of High Temperature Plasma to very low X-ray spectroscopy. The main purpose of this detector is to provide a state-of-the-art solution to upgrade the current detector system of the NPAs at JET and also to pave the road for the future detection systems of the ITER experimental reactor. Currently the NPAs use a very thin scintillator-photomultiplier tube [F. García, S.S. Kozlovsky, D.V. Balin, Background Properties of CEM, MCP and PMT detectors at n-γ irradiation. Preprint PNPI-2392, Gatchina, 2000, p. 9 [1]; F. García, S.S. Kozlovsky, V.V. Ianovsky, Scintillation Detectors with Low Sensitivity to n-γ Background. Preprint PNPI-2391, Gatchina, 2000, p. 8 [2

  1. Preliminary RAMI analysis of DFLL TBS for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dagui [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230031 (China); Yuan, Run [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Jiaqun, E-mail: jiaqun.wang@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wang, Fang; Wang, Jin [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

    2016-11-15

    Highlights: • We performed the functional analysis of the DFLL TBS. • We performed a failure mode analysis of the DFLL TBS. • We estimated the reliability and availability of the DFLL TBS. • The ITER RAMI approach was applied to the DFLL TBS for technical risk control in the design phase. - Abstract: ITER is the first fusion machine fully designed to prove the physics and technological basis for next fusion power plants. Among the main technical objectives of ITER is to test and validate design concepts of tritium breeding blankets relevant to the fusion power plants. To achieve this goal, China has proposed the dual functional lithium-lead test blanket module (DFLL TBM) concept design. The DFLL TBM and its associated ancillary system were called DFLL TBS. The DFLL TBS play a key role in next fusion reactor. In order to ensure reliable and available of DFLL TBS, the risk control project of DFLL TBS has been put on the schedule. As the stage of the ITER technical risk control policy, the RAMI (Reliability, Availability, Maintainability, Inspectability) approach was used to control the technical risk of ITER. In this paper, the RAMI approach was performed on the conceptual design of DFLL TBS. A functional breakdown was prepared on DFLL TBS, and the system was divided into 3 main functions and 72 basic functions. Based on the result of functional breakdown of DFLL TBS, the reliability block diagrams were prepared to estimate the reliability and availability of each function under the stipulated operating conditions. The inherent availability of the DFLL TBS expected after implementation of mitigation actions was calculated to be 98.57% over 2 years based on the ITER reliability database. A Failure Modes Effects and Criticality Analysis (FMECA) was performed with criticality charts highlighting the risk level of the different failure modes with regard to their probability of occurrence and their effects on the availability.

  2. Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

    NARCIS (Netherlands)

    Banerjee, S.; Vasu, P.; von Hellermann, M.; Jaspers, R. J. E.

    2010-01-01

    Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is,

  3. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  4. NEUTRONIC REACTORS

    Science.gov (United States)

    Wigner, E.P.; Young, G.J.

    1958-10-14

    A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

  5. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  6. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. (eds.)

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  7. Accuracy improvement of a hybrid robot for ITER application using POE modeling method

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yongbo, E-mail: yongbo.wang@hotmail.com [Laboratory of Intelligent Machines, Lappeenranta University of Technology, FIN-53851 Lappeenranta (Finland); Wu, Huapeng; Handroos, Heikki [Laboratory of Intelligent Machines, Lappeenranta University of Technology, FIN-53851 Lappeenranta (Finland)

    2013-10-15

    Highlights: ► The product of exponential (POE) formula for error modeling of hybrid robot. ► Differential Evolution (DE) algorithm for parameter identification. ► Simulation results are given to verify the effectiveness of the method. -- Abstract: This paper focuses on the kinematic calibration of a 10 degree-of-freedom (DOF) redundant serial–parallel hybrid robot to improve its accuracy. The robot was designed to perform the assembling and repairing tasks of the vacuum vessel (VV) of the international thermonuclear experimental reactor (ITER). By employing the product of exponentials (POEs) formula, we extended the POE-based calibration method from serial robot to redundant serial–parallel hybrid robot. The proposed method combines the forward and inverse kinematics together to formulate a hybrid calibration method for serial–parallel hybrid robot. Because of the high nonlinear characteristics of the error model and too many error parameters need to be identified, the traditional iterative linear least-square algorithms cannot be used to identify the parameter errors. This paper employs a global optimization algorithm, Differential Evolution (DE), to identify parameter errors by solving the inverse kinematics of the hybrid robot. Furthermore, after the parameter errors were identified, the DE algorithm was adopted to numerically solve the forward kinematics of the hybrid robot to demonstrate the accuracy improvement of the end-effector. Numerical simulations were carried out by generating random parameter errors at the allowed tolerance limit and generating a number of configuration poses in the robot workspace. Simulation of the real experimental conditions shows that the accuracy of the end-effector can be improved to the same precision level of the given external measurement device.

  8. Iterative reconstruction of volumetric particle distribution

    Science.gov (United States)

    Wieneke, Bernhard

    2013-02-01

    For tracking the motion of illuminated particles in space and time several volumetric flow measurement techniques are available like 3D-particle tracking velocimetry (3D-PTV) recording images from typically three to four viewing directions. For higher seeding densities and the same experimental setup, tomographic PIV (Tomo-PIV) reconstructs voxel intensities using an iterative tomographic reconstruction algorithm (e.g. multiplicative algebraic reconstruction technique, MART) followed by cross-correlation of sub-volumes computing instantaneous 3D flow fields on a regular grid. A novel hybrid algorithm is proposed here that similar to MART iteratively reconstructs 3D-particle locations by comparing the recorded images with the projections calculated from the particle distribution in the volume. But like 3D-PTV, particles are represented by 3D-positions instead of voxel-based intensity blobs as in MART. Detailed knowledge of the optical transfer function and the particle image shape is mandatory, which may differ for different positions in the volume and for each camera. Using synthetic data it is shown that this method is capable of reconstructing densely seeded flows up to about 0.05 ppp with similar accuracy as Tomo-PIV. Finally the method is validated with experimental data.

  9. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  10. The ITER project technological challenges

    CERN Document Server

    CERN. Geneva; Lister, Joseph; Marquina, Miguel A; Todesco, Ezio

    2005-01-01

    The first lecture reminds us of the ITER challenges, presents hard engineering problems, typically due to mechanical forces and thermal loads and identifies where the physics uncertainties play a significant role in the engineering requirements. The second lecture presents soft engineering problems of measuring the plasma parameters, feedback control of the plasma and handling the physics data flow and slow controls data flow from a large experiment like ITER. The last three lectures focus on superconductors for fusion. The third lecture reviews the design criteria and manufacturing methods for 6 milestone-conductors of large fusion devices (T-7, T-15, Tore Supra, LHD, W-7X, ITER). The evolution of the designer approach and the available technologies are critically discussed. The fourth lecture is devoted to the issue of performance prediction, from a superconducting wire to a large size conductor. The role of scaling laws, self-field, current distribution, voltage-current characteristic and transposition are...

  11. ITER Disruption Mitigation System Design

    Science.gov (United States)

    Rasmussen, David; Lyttle, M. S.; Baylor, L. R.; Carmichael, J. R.; Caughman, J. B. O.; Combs, S. K.; Ericson, N. M.; Bull-Ezell, N. D.; Fehling, D. T.; Fisher, P. W.; Foust, C. R.; Ha, T.; Meitner, S. J.; Nycz, A.; Shoulders, J. M.; Smith, S. F.; Warmack, R. J.; Coburn, J. D.; Gebhart, T. E.; Fisher, J. T.; Reed, J. R.; Younkin, T. R.

    2015-11-01

    The disruption mitigation system for ITER is under design and will require injection of up to 10 kPa-m3 of deuterium, helium, neon, or argon material for thermal mitigation and up to 100 kPa-m3 of material for suppression of runaway electrons. A hybrid unit compatible with the ITER nuclear, thermal and magnetic field environment is being developed. The unit incorporates a fast gas valve for massive gas injection (MGI) and a shattered pellet injector (SPI) to inject a massive spray of small particles, and can be operated as an SPI with a frozen pellet or an MGI without a pellet. Three ITER upper port locations will have three SPI/MGI units with a common delivery tube. One equatorial port location has space for sixteen similar SPI/MGI units. Supported by US DOE under DE-AC05-00OR22725.

  12. Neutronics calculation of an heterogeneous compact and thermal core by means of deterministic and stochastic transport theory. Application to the experimental reactor of the University of Strasbourg; Modelisation neutronique d`un coeur thermique compact et heterogene en theorie du transport deterministe et probabiliste. Application au reacteur experimental de l`Universite de Strasbourg

    Energy Technology Data Exchange (ETDEWEB)

    Jammes, Ch

    1997-11-28

    The aim of this work is to create, validate theoretically and experimentally a calculation route for a thermal irradiation reactor. This is the research reactor of the University of Strasbourg, which presents all of characteristics of this reactor-type: compact and heterogeneous core, slab-type fuel with a high 235-uranium enrichment. This calculation route is based on the first use of the following two modern transport methods: the TDT method and the Monte Carlo method. The former, programmed within the APOLLO2 code, is a two dimensional collision probabilities method. The later, used by the TRIPOLI4 code, is a stochastic method. Both can be applied to complex geometries. After a few theoretical reminders about transport codes, a set of integral experiments is described which have been realized within the research reactor of the University of Strasbourg. One of them has been performed for this study. At the beginning of the theoretical part, significant errors are apparent due to the use of calculation route based on homogenization, condensation and the diffusion approximation. An extensive comparison between the discrete ordinates method and the TDT method carries out that the use of the TDT method is relevant for the studied reactor. The treatment of axial leakage with this method is the only disadvantage. Therefore, the use of the code TRIPOLI4 is recommended for a more accurate study of leakage within a reflector. By means of the experimental data, the ability of our calculation route is confirmed for essential neutronics questions such as the critical mass determination, the power distribution and the fuel management. (author)

  13. Experimental studies into the fluid dynamic performance of the coolant flow in the mixed core of the Temelin NPP VVER-1000 reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-11-01

    Full Text Available The paper presents the results of studies into the interassembly coolant interaction in the Temelin nuclear power plant (NPP VVER-1000 reactor core. An aerodynamic test bench was used to study the coolant flow processes in a TVSA-type fuel assembly bundle. To obtain more detailed information on the coolant flow dynamics, a VVER-1000 reactor core fragment was selected as the test model, which comprised two segments of a TVSA-12 PLUS fuel assembly and one segment of a TVSA-T assembly with stiffening angles and an interassembly gap. The studies into the coolant fluid dynamics consisted in measuring the velocity vector both in representative TVSA regions and inside the interassembly gap using a five-channel pneumometric probe. An analysis into the spatial distribution of the absolute flow velocity projections made it possible to detail the TVSA spacer, mixing and combined spacer grid flow pattern, identify the regions with the maximum transverse coolant flow, and determine the depth of the coolant flow disturbance propagation and redistribution in adjacent TVSA assemblies. The results of the studies into the interassembly coolant interaction among the adjacent TVSA assemblies are used at OKBM Afrikantov to update the VVER-1000 core thermal-hydraulic analysis procedures and have been added to the database for verification of computational fluid dynamics (CFD codes and for detailed cellwise analyses of the VVER-100 reactor cores.

  14. ITER CS Model Coil and CS Insert Test Results

    Energy Technology Data Exchange (ETDEWEB)

    Martovetsky, N; Michael, P; Minervina, J; Radovinsky, A; Takayasu, M; Thome, R; Ando, T; Isono, T; Kato, T; Nakajima, H; Nishijima, G; Nunoya, Y; Sugimoto, M; Takahashi, Y; Tsuji, H; Bessette, D; Okuno, K; Ricci, M

    2000-09-07

    The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the main results of the tests performed.

  15. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

  16. Neutronic reactor

    Science.gov (United States)

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  17. Neutronic reactor

    Energy Technology Data Exchange (ETDEWEB)

    Babcock, D.F.; Menegus, R.L.; Wende, C.W.

    1983-01-04

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  18. On the Short Horizon of Spontaneous Iterative Reasoning in Logical Puzzles and Games

    Science.gov (United States)

    Mazzocco, Ketti; Cherubini, Anna Maria; Cherubini, Paolo

    2013-01-01

    A reasoning strategy is iterative when the initial conclusion suggested by a set of premises is integrated into that set of premises in order to yield additional conclusions. Previous experimental studies on game theory-based strategic games (such as the beauty contest game) observed difficulty in reasoning iteratively, which has been partly…

  19. Turbulent precipitation of uranium oxalate in a vortex reactor - experimental study and modelling; Precipitation turbulente d'oxalate d'uranium en reacteur vortex - etude experimentale et modelisation

    Energy Technology Data Exchange (ETDEWEB)

    Sommer de Gelicourt, Y

    2004-03-15

    Industrial oxalic precipitation processed in an un-baffled magnetically stirred tank, the Vortex Reactor, has been studied with uranium simulating plutonium. Modelling precipitation requires a mixing model for the continuous liquid phase and the solution of population balance for the dispersed solid phase. Being chemical reaction influenced by the degree of mixing at molecular scale, that commercial CFD code does not resolve, a sub-grid scale model has been introduced: the finite mode probability density functions, and coupled with a model for the liquid energy spectrum. Evolution of the dispersed phase has been resolved by the quadrature method of moments, first used here with experimental nucleation and growth kinetics, and an aggregation kernel based on local shear rate. The promising abilities of this local approach, without any fitting constant, are strengthened by the similarity between experimental results and simulations. (author)

  20. Dynamic model of biological nitrogen removal in SBR reactors: Experimental study and model calibration; Modello di simulazione dinamica della rimozione biologica dell`azoto con reattori a flusso discontinuo (SBR): indagine sperimentale e taratura del modello

    Energy Technology Data Exchange (ETDEWEB)

    Andreottola, G. [Trento Univ. (Italy). Dipt. di Ingegneria Civile e Ambientale; Bortone, G.; Spadoni, S. [ENEA, Bologna (Italy). Area Energia Ambiente e Salute

    1995-03-01

    In the present study are presented the results of the experimental activity carried out in order to calibrate and verify the dynamical model developed by the Authors for nitrogen removal with SBR processes. Four track studies (two for the calibration phase and two for the verification one) have been carried out on the SBR lab-scale reactor. The experimental results have shown temporary increase of nitrite in the nitrification phase. Therefore, the SBR dynamic model has been modified, in order to better explain the dynamic behavior of the nitrogen removal process: nitrousation and nitratation processes have been explicitly described and a inhibition kinetic of nitratation by ammonia has been introduced. The new model, calibrated and verified, has successfully explained the dynamic behavior of the process. Further work will be carried out in order to better understand the inhibition dynamic of nitratation.

  1. Optimization of Fast Critical Experiments to Reduce Nuclear Data Uncertainties in Support of a Fast Burner Reactor Design Concept

    Science.gov (United States)

    Stover, Tracy E., Jr.

    An optimization technique has been developed to select optimized experimental design specifications to produce data specifically designed to be assimilated to optimize a given reactor concept. Data from the optimized experiment is assimilated to generate posteriori uncertainties on the reactor concept's core attributes from which the design responses are computed. The reactor concept is then optimized with the new data to realize cost savings by reducing margin. The optimization problem iterates until an optimal experiment is found to maximize the savings. A new generation of innovative nuclear reactor designs, in particular fast neutron spectrum recycle reactors, are being considered for the application of closing the nuclear fuel cycle in the future. Safe and economical design of these reactors will require uncertainty reduction in basic nuclear data which are input to the reactor design. These data uncertainty propagate to design responses which in turn require the reactor designer to incorporate additional safety margin into the design, which often increases the cost of the reactor. Therefore basic nuclear data needs to be improved and this is accomplished through experimentation. Considering the high cost of nuclear experiments, it is desired to have an optimized experiment which will provide the data needed for uncertainty reduction such that a reactor design concept can meet its target accuracies or to allow savings to be realized by reducing the margin required due to uncertainty propagated from basic nuclear data. However, this optimization is coupled to the reactor design itself because with improved data the reactor concept can be re-optimized itself. It is thus desired to find the experiment that gives the best optimized reactor design. Methods are first established to model both the reactor concept and the experiment and to efficiently propagate the basic nuclear data uncertainty through these models to outputs. The representativity of the experiment

  2. Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors

    Science.gov (United States)

    Lay, Wei-Siang

    Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled

  3. Design and first applications of the ITER integrated modelling & analysis suite

    Science.gov (United States)

    Imbeaux, F.; Pinches, S. D.; Lister, J. B.; Buravand, Y.; Casper, T.; Duval, B.; Guillerminet, B.; Hosokawa, M.; Houlberg, W.; Huynh, P.; Kim, S. H.; Manduchi, G.; Owsiak, M.; Palak, B.; Plociennik, M.; Rouault, G.; Sauter, O.; Strand, P.

    2015-11-01

    The ITER Integrated Modelling & Analysis Suite (IMAS) will support both plasma operation and research activities on the ITER tokamak experiment. The IMAS will be accessible to all ITER members as a key tool for the scientific exploitation of ITER. The backbone of the IMAS infrastructure is a standardized, machine-generic data model that represents simulated and experimental data with identical structures. The other outcomes of the IMAS design and prototyping phase are a set of tools to access data and design integrated modelling workflows, as well as first plasma simulators workflows and components implemented with various degrees of modularity.

  4. Baseline Architecture of ITER Control System

    Science.gov (United States)

    Wallander, A.; Di Maio, F.; Journeaux, J.-Y.; Klotz, W.-D.; Makijarvi, P.; Yonekawa, I.

    2011-08-01

    The control system of ITER consists of thousands of computers processing hundreds of thousands of signals. The control system, being the primary tool for operating the machine, shall integrate, control and coordinate all these computers and signals and allow a limited number of staff to operate the machine from a central location with minimum human intervention. The primary functions of the ITER control system are plant control, supervision and coordination, both during experimental pulses and 24/7 continuous operation. The former can be split in three phases; preparation of the experiment by defining all parameters; executing the experiment including distributed feed-back control and finally collecting, archiving, analyzing and presenting all data produced by the experiment. We define the control system as a set of hardware and software components with well defined characteristics. The architecture addresses the organization of these components and their relationship to each other. We distinguish between physical and functional architecture, where the former defines the physical connections and the latter the data flow between components. In this paper, we identify the ITER control system based on the plant breakdown structure. Then, the control system is partitioned into a workable set of bounded subsystems. This partition considers at the same time the completeness and the integration of the subsystems. The components making up subsystems are identified and defined, a naming convention is introduced and the physical networks defined. Special attention is given to timing and real-time communication for distributed control. Finally we discuss baseline technologies for implementing the proposed architecture based on analysis, market surveys, prototyping and benchmarking carried out during the last year.

  5. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  6. Iterated Dynamic Condensation Technique and Its Applications in Modal Testing

    Directory of Open Access Journals (Sweden)

    Zhikun Hou

    1997-01-01

    Full Text Available This article addresses an application of the dynamic condensation in modal testing, which requires measurements of natural frequencies and a limited number of components of modal shapes only. The measured frequencies can be employed as the initial estimates in the iterated dynamic condensation to effectively find an exact solution for the original eigenvalue problem and determine the transformation matrix between the testing and nontesting coordinates. Complete modal shapes including the nontesting coordinates can then be estimated based on the limited measurements. Numerical results are presented to illustrate the feasibility and efficiency of the iterated dynamic condensation. A comparison with the experimental results of a multiple disk system is also provided.

  7. Iterative reconstruction of the detector response for medical gamma cameras

    CERN Document Server

    Morozov, A; Alves, F; Domingos, V; Martins, R; Neves, F; Chepel, V

    2015-01-01

    Statistical event reconstruction techniques can give better results for gamma cameras than the traditional centroid method. However, implementation of such techniques requires detailed knowledge of the PMT light response functions. Here we describe an iterative technique which allows to obtain the response functions from flood irradiation data without imposing strict requirements on the spatial uniformity of the event distribution. A successful application of the technique for medical gamma cameras is demonstrated using both simulated and experimental data. We show that this technique can be used for monitoring of the photomultiplier gain variations. An implementation of the iterative reconstruction technique capable of operating in real-time is also presented.

  8. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  9. Jules Horowitz Reactor, basic design

    Energy Technology Data Exchange (ETDEWEB)

    Bergamaschi, Y.; Bouilloux, Y.; Chantoin, P.; Guigon, B.; Bravo, X.; Germain, C.; Rommens, M.; Tremodeux, P

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: - represent a significant step in term of performances and experimental capabilities, - be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements, - reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-27

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

  12. Experimental and density functional theoretical study of the effects of Fenton’s reaction on the degradation of Bisphenol A in a high voltage plasma reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Fei, E-mail: daif@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Fan, Xiangru, E-mail: fanx@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Stratton, Gunnar R., E-mail: strattgr@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Bellona, Christopher L., E-mail: cbellona@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Department of Civil and Environmental Engineering, 1500 Illinois St., Colorado School of Mines, Golden, 80401 CO (United States); Holsen, Thomas M., E-mail: tholsen@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Crimmins, Bernard S., E-mail: bcrimmin@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Xia, Xiaoyan, E-mail: xiax@clarkson.edu [Department of Civil and Environmental Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States); Mededovic Thagard, Selma, E-mail: smededov@clarkson.edu [Department of Chemical and Biomolecular Engineering, 8 Clarkson Avenue, Clarkson University, Potsdam, 13699 NY (United States)

    2016-05-05

    Highlights: • Combining the Fenton reaction with the plasma treatment reduces Bisphenol A concentration below the detection limit within 30 min. • Carbon steel electrode in the plasma reactor can be used as a source of iron ions. • OH radical attack on Bisphenol A is the primary pathway for byproduct formation. - Abstract: A novel electrical discharge plasma reactor configuration with and without iron ions was evaluated for the degradation of 0.02 mM Bisphenol A (BPA). The pseudo-first-order reaction rate constant calculated for the plasma treatment of BPA with a stainless steel electrode in the presence of dissolved ferrous ion (Fe{sup 2+}) salts (termed plasma/Fenton treatment) was higher than in the plasma treatment in the absence of iron salts. At the optimal ferrous ion concentration, longer plasma treatment times resulted in higher BPA degradation rates, likely due to increased hydroxyl (OH) radical concentration formed through the decomposition of H{sub 2}O{sub 2}. Replacing the stainless steel with a carbon steel grounded electrode resulted in the release of iron ions from the carbon steel thereby increasing the rate of BPA removal and eliminating the need for iron salts. After the plasma/Fenton treatment, >97% of the residual iron salts were removed by coagulation/flocculation/sedimentation. Byproduct identification coupled with density functional theory (DFT) calculations confirmed that OH radical attack on BPA’s hydroxyl group is the primary pathway for byproduct formation.

  13. Mechanical properties of type 316L stainless steel welded joint for ITER vacuum vessel (1). Experiment of unirradiated welded joint

    Energy Technology Data Exchange (ETDEWEB)

    Saito, Shigeru; Fukaya, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishiyama, Shintaro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takahashi, Hiroyuki; Koizumi, Kouichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-01-01

    In design activity of ITER, the vacuum vessel (VV) is ranked as one of the most important components in core reactor from the view point of first barrier to tritium release from the reactor. The VV of ITER is designed as double walled structure so that some parts of them are not qualified in the conventional design standards. So it is necessary to prepare the new design standards to be applied them. JAERI has executed the preparation activity of the new design standards and the technical data to support them. In this study, the results of metallographic observation and mechanical properties of unirradiated type 316L stainless steel welded joint were reported. (author)

  14. Diverse Power Iteration Embeddings and Its Applications

    Energy Technology Data Exchange (ETDEWEB)

    Huang H.; Yoo S.; Yu, D.; Qin, H.

    2014-12-14

    Abstract—Spectral Embedding is one of the most effective dimension reduction algorithms in data mining. However, its computation complexity has to be mitigated in order to apply it for real-world large scale data analysis. Many researches have been focusing on developing approximate spectral embeddings which are more efficient, but meanwhile far less effective. This paper proposes Diverse Power Iteration Embeddings (DPIE), which not only retains the similar efficiency of power iteration methods but also produces a series of diverse and more effective embedding vectors. We test this novel method by applying it to various data mining applications (e.g. clustering, anomaly detection and feature selection) and evaluating their performance improvements. The experimental results show our proposed DPIE is more effective than popular spectral approximation methods, and obtains the similar quality of classic spectral embedding derived from eigen-decompositions. Moreover it is extremely fast on big data applications. For example in terms of clustering result, DPIE achieves as good as 95% of classic spectral clustering on the complex datasets but 4000+ times faster in limited memory environment.

  15. Iterative Learning Control with Forgetting Factor for Urban Road Network

    Directory of Open Access Journals (Sweden)

    Tianyi Lan

    2017-01-01

    Full Text Available In order to improve the traffic condition, a novel iterative learning control (ILC algorithm with forgetting factor for urban road network is proposed by using the repeat characteristics of traffic flow in this paper. Rigorous analysis shows that the proposed ILC algorithm can guarantee the asymptotic convergence. Through iterative learning control of the traffic signals, the number of vehicles on each road in the network can gradually approach the desired level, thereby preventing oversaturation and traffic congestion. The introduced forgetting factor can effectively adjust the control input according to the states of the system and filter along the direction of the iteration. The results show that the forgetting factor has an important effect on the robustness of the system. The theoretical analysis and experimental simulations are given to verify the validity of the proposed method.

  16. Key ITER plasma edge and plasma-material interaction issues

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. E-mail: federig@ipp.mpg.de; Andrew, P.; Barabaschi, P.; Brooks, J.; Doerner, R.; Geier, A.; Herrmann, A.; Janeschitz, G.; Krieger, K.; Kukushkin, A.; Loarte, A.; Neu, R.; Saibene, G.; Shimada, M.; Strohmayer, G.; Sugihara, M

    2003-03-01

    Some of the remaining crucial plasma edge physics and plasma-material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

  17. Iterative Learning Impedance for Lower Limb Rehabilitation Robot.

    Science.gov (United States)

    Guo, Chenhui; Guo, Shuai; Ji, Jiancheng; Xi, Fengfeng

    2017-01-01

    This paper discusses the problem of squatting training of stroke patients. The main idea is to correct the patient's training trajectory through an iterative learning control (ILC) method. To obtain better rehabilitation effect, a patient will typically be required to practice a reference posture for many times, while most of active training methods can hardly keep the patients training with correct posture. Instead of the conventional ILC strategy, an impedance-based iterative learning method is proposed to regulate the impedance value dynamically and smartly which will help patients correct their posture gradually and perform better. To facilitate impedance-based ILC, we propose two objectives. The first objective is to find the suitable values of impedance based on the ILC scheme. The second objective is to search the moderate learning convergence speed and robustness in the iterative domain. The simulation and experimental results demonstrate that the performance of trajectory tracking will be improved greatly via the proposed algorithm.

  18. Iterative Perceptual Learning for Social Behavior Synthesis

    NARCIS (Netherlands)

    de Kok, I.A.; Poppe, Ronald Walter; Heylen, Dirk K.J.

    We introduce Iterative Perceptual Learning (IPL), a novel approach for learning computational models for social behavior synthesis from corpora of human-human interactions. The IPL approach combines perceptual evaluation with iterative model refinement. Human observers rate the appropriateness of

  19. Existence test for asynchronous interval iterations

    DEFF Research Database (Denmark)

    Madsen, Kaj; Caprani, O.; Stauning, Ole

    1997-01-01

    In the search for regions that contain fixed points ofa real function of several variables, tests based on interval calculationscan be used to establish existence ornon-existence of fixed points in regions that are examined in the course ofthe search. The search can e.g. be performed...... as a synchronous (sequential) interval iteration:In each iteration step all components of the iterate are calculatedbased on the previous iterate. In this case it is straight forward to base simple interval existence and non-existencetests on the calculations done in each step of the iteration. The search can also...... be performed as an asynchronous (parallel) iteration: Only a few components are changed in each stepand this calculation is in general based on components from differentprevious iterates. For the asynchronous iteration it turns out thatsimple tests of existence and non-existence can be based...

  20. Iterative perceptual learning for social behavior synthesis

    NARCIS (Netherlands)

    de Kok, I.A.; Poppe, Ronald Walter; Heylen, Dirk K.J.

    We introduce Iterative Perceptual Learning (IPL), a novel approach to learn computational models for social behavior synthesis from corpora of human–human interactions. IPL combines perceptual evaluation with iterative model refinement. Human observers rate the appropriateness of synthesized

  1. Intelligent controller of a flexible hybrid robot machine for ITER assembly and maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Al-saedi, Mazin I., E-mail: mazin.al-saedi@lut.fi; Wu, Huapeng; Handroos, Heikki

    2014-10-15

    Highlights: • Studying flexible multibody dynamic of hybrid parallel robot. • Investigating fuzzy-PD controller to control a hybrid flexible hydraulically driven robot. • Investigating ANFIS-PD controller to control a hybrid flexible robot. Compare to traditional PID this method gives better performance. • Using the equilibrium of reaction forces between the parallel and serial parts of hybrid robot to control the serial part hydraulically driven. - Abstract: The assembly and maintenance of International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is highly challenging since the tasks performed by the robot involve welding, material handling, and machine cutting from inside the VV. To fulfill the tasks in ITER application, this paper presents a hybrid redundant manipulator with four DOFs provided by serial kinematic axes and six DOFs by parallel mechanism. Thus, in machining, to achieve greater end-effector trajectory tracking accuracy for surface quality, a robust control of the actuators for the flexible link has to be deduced. In this paper, the intelligent control of a hydraulically driven parallel robot part based on the dynamic model and two control schemes have been investigated: (1) fuzzy-PID self tuning controller composed of the conventional PID control and with fuzzy logic; (2) adaptive neuro-fuzzy inference system-PID (ANFIS-PID) self tuning of the gains of the PID controller, which are implemented independently to control each hydraulic cylinder of the parallel robot based on rod position predictions. The obtained results of the fuzzy-PID and ANFIS-PID self tuning controller can reduce more tracking errors than the conventional PID controller. Subsequently, the serial component of the hybrid robot can be analyzed using the equilibrium of reaction forces at the universal joint connections of the hexa-element. To achieve precise positional control of the end effector for maximum precision machining, the hydraulic cylinder should

  2. Monte Carlo Alpha Iteration Algorithm for a Subcritical System Analysis

    Directory of Open Access Journals (Sweden)

    Hyung Jin Shim

    2015-01-01

    Full Text Available The α-k iteration method which searches the fundamental mode alpha-eigenvalue via iterative updates of the fission source distribution has been successfully used for the Monte Carlo (MC alpha-static calculations of supercritical systems. However, the α-k iteration method for the deep subcritical system analysis suffers from a gigantic number of neutron generations or a huge neutron weight, which leads to an abnormal termination of the MC calculations. In order to stably estimate the prompt neutron decay constant (α of prompt subcritical systems regardless of subcriticality, we propose a new MC alpha-static calculation method named as the α iteration algorithm. The new method is derived by directly applying the power method for the α-mode eigenvalue equation and its calculation stability is achieved by controlling the number of time source neutrons which are generated in proportion to α divided by neutron speed in MC neutron transport simulations. The effectiveness of the α iteration algorithm is demonstrated for two-group homogeneous problems with varying the subcriticality by comparisons with analytic solutions. The applicability of the proposed method is evaluated for an experimental benchmark of the thorium-loaded accelerator-driven system.

  3. SEU mitigation exploratory tests in a ITER related FPGA

    Energy Technology Data Exchange (ETDEWEB)

    Batista, Antonio J.N., E-mail: toquim@ipfn.tecnico.ulisboa.pt [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Leong, Carlos [Instituto de Engenharia de Sistemas e Computadores – Investigação e Desenvolvimento (INESC-ID), 1000-029 Lisboa (Portugal); Santos, Bruno; Fernandes, Ana [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Ramos, Ana Rita; Santos, Joana P.; Marques, José G. [Centro de Ciências e Tecnologias Nucleares (C2TN), Instituto Superior Técnico (IST), Universidade de Lisboa - UL, 2695-066 Bobadela (Portugal); Teixeira, Isabel C.; Teixeira, João P. [Instituto de Engenharia de Sistemas e Computadores – Investigação e Desenvolvimento (INESC-ID), 1000-029 Lisboa (Portugal); Sousa, Jorge; Gonçalves, Bruno [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal)

    2017-05-15

    Data acquisition hardware of ITER diagnostics if located in the port cells of the tokamak, as an example, will be irradiated with neutrons during the fusion reactor operation. Due to this reason the majority of the hardware containing Field Programmable Gate Arrays (FPGA) will be placed after the ITER bio-shield, such as the cubicles instrumentation room. Nevertheless, it is worth to explore real-time mitigation of soft-errors caused by neutrons radiation in ITER related FPGAs. A Virtex-6 FPGA from Xilinx (XC6VLX365T-1FFG1156C) is used on the ATCA-IO-PROCESSOR board, included in the ITER Catalog of Instrumentation & Control (I & C) products – Fast Controllers. The Virtex-6 is a re-programmable logic device where the configuration is stored in Static RAM (SRAM), the functional data is stored in dedicated Block RAM (BRAM) and the functional state logic in Flip-Flops. Single Event Upsets (SEU) due to the ionizing radiation of neutrons cause soft errors, unintended changes (bit-flips) of the logic values stored in the state elements of the FPGA. Real-time SEU monitoring and soft errors repairing, when possible, were explored in this work. An FPGA built-in Soft Error Mitigation (SEM) controller detects and corrects soft errors in the FPGA Configuration Memory (CM). BRAM based SEU sensors with Error Correction Code (ECC) detect and repair the respective BRAM contents. Real-time mitigation of SEU can increase reliability and availability of data acquisition hardware for nuclear applications. The results of the tests performed using the SEM controller and the SEU sensors are presented for a Virtex-6 FPGA (XC6VLX240T-1FFG1156C) when irradiated with neutrons from the Portuguese Research Reactor (RPI), a 1 MW nuclear fission reactor, operated by IST in the neighborhood of Lisbon. Results show that the proposed SEU mitigation technique is able to repair the majority of the detected SEU soft-errors in the FPGA memory.

  4. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  5. Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Leininger, Simon

    2017-02-22

    In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

  6. Distributed Video Coding: Iterative Improvements

    DEFF Research Database (Denmark)

    Luong, Huynh Van

    at the decoder side offering such benefits for these applications. Although there have been some advanced improvement techniques, improving the DVC coding efficiency is still challenging. The thesis addresses this challenge by proposing several iterative algorithms at different working levels, e.g. bitplane......, band, and frame levels. In order to show the information theoretic basis, theoretical foundations of DVC are introduced. The first proposed algorithm applies parallel iterative decoding using multiple LDPC decoders to utilize cross bitplane correlation. To improve Side Information (SI) generation...... and noise modeling and also learn from the previous decoded Wyner-Ziv (WZ) frames, side information and noise learning (SING) is proposed. The SING scheme introduces an optical flow technique to compensate the weaknesses of the block based SI generation and also utilizes clustering of DCT blocks to capture...

  7. Nuclear Reactors. Revised.

    Science.gov (United States)

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  8. Modeling Results For the ITER Cryogenic Fore Pump. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Pfotenhauer, John M. [University of Wisconsin, Madison, WI (United States); Zhang, Dongsheng [University of Wisconsin, Madison, WI (United States)

    2014-03-31

    A numerical model characterizing the operation of a cryogenic fore-pump (CFP) for ITER has been developed at the University of Wisconsin – Madison during the period from March 15, 2011 through June 30, 2014. The purpose of the ITER-CFP is to separate hydrogen isotopes from helium gas, both making up the exhaust components from the ITER reactor. The model explicitly determines the amount of hydrogen that is captured by the supercritical-helium-cooled pump as a function of the inlet temperature of the supercritical helium, its flow rate, and the inlet conditions of the hydrogen gas flow. Furthermore the model computes the location and amount of hydrogen captured in the pump as a function of time. Throughout the model’s development, and as a calibration check for its results, it has been extensively compared with the measurements of a CFP prototype tested at Oak Ridge National Lab. The results of the model demonstrate that the quantity of captured hydrogen is very sensitive to the inlet temperature of the helium coolant on the outside of the cryopump. Furthermore, the model can be utilized to refine those tests, and suggests methods that could be incorporated in the testing to enhance the usefulness of the measured data.

  9. Decay heat experiment and validation of calculation code systems for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

    1999-10-01

    Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

  10. Mahalanobis Distance Based Iterative Closest Point

    DEFF Research Database (Denmark)

    Hansen, Mads Fogtmann; Blas, Morten Rufus; Larsen, Rasmus

    2007-01-01

    a fast optimization. Initially, the covariance matrices are set to the identity matrix, and all shapes are aligned to a randomly selected shape (equivalent to standard ICP). From this point the algorithm iterates between the steps: (a) obtain mean shape and new estimates of the covariance matrices from...... the notion of a mahalanobis distance map upon a point set with associated covariance matrices which in addition to providing correlation weighted distance implicitly provides a method for assigning correspondence during alignment. This distance map provides an easy formulation of the ICP problem that permits...... the aligned shapes, (b) align shapes to the mean shape. Three different methods for estimating the mean shape with associated covariance matrices are explored in the paper. The proposed methods are validated experimentally on two separate datasets (IMM face dataset and femur-bones). The superiority of ICP...

  11. Multiphase Reactors: Models and Experimental Verification Les réacteurs polyphasés : modèles et vérification expérimentale

    Directory of Open Access Journals (Sweden)

    Dudukovic M. P.

    2006-11-01

    Full Text Available Identification and quantification of flow regimes, phase holdup distributions, flow patterns and backmixing is essential for proper design and scale-up of multiphase reactors. Existing models often suffer from inadequate experimental confirmation. Here, we describe recent progress made in our laboratory in characterizing liquid circulation and backmixing in bubble columns and in quantifying, via a phenomenological model, the behavior of trickle-beds in the low interaction regime. The need for imaging techniques such as fan-beam tomography and PET is illustrated. Liquid mixing in bubble columns caused by global liquid convection and by turbulent eddies is usually described by the axial dispersion model (ADM which, at least in the churn turbulent regime, lacks a theoretical basis. In the ADM the two mixing mechanisms are lumped into a single effective dispersion coefficient. An alternative backmixing description assumes multiple liquid circulation cells, with height equal to column diameter, to occur in the column. We have successfully demonstrated that instantaneous and time averaged liquid velocity profiles in the column can be determined by monitoring the motion of a neutrally bouyant tracer particle via a Computer Aided Radioactive Particle Tracking (CARPT technique. It can now be shown that a single strong liquid circulation cell exists in columns of a variety of diameters and at different operating conditions but that smaller circulation cells can form in the distributor region. Model predicted one dimensional axial time averaged liquid velocity profile agrees well with the data and applies to the middle section of the column. The CARPT technique allows the computation of Lagrangian autocorrelation coefficients, the rms fluctuating velocities, Lagrangian integral time scales and the turbulent dispersion coefficients. Thus CARPT provides the necessary information i. e. velocity profiles and turbulent dispersion coefficients, for

  12. Photocatalytic reactor

    Science.gov (United States)

    Bischoff, Brian L.; Fain, Douglas E.; Stockdale, John A. D.

    1999-01-01

    A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

  13. Counteracting 16-QAM Optical Fiber Transmission Impairments With Iterative Turbo Equalization

    DEFF Research Database (Denmark)

    Arlunno, Valeria; Caballero Jambrina, Antonio; Borkowski, Robert

    2013-01-01

    A turbo equalization (TE) scheme based on convolutional code and normalized least mean square equalizer for coherent optical communication links is proposed and experimentally demonstrated. The proposed iterative TE technique is proved effective for counteracting polarization-division-multiplexin......A turbo equalization (TE) scheme based on convolutional code and normalized least mean square equalizer for coherent optical communication links is proposed and experimentally demonstrated. The proposed iterative TE technique is proved effective for counteracting polarization...

  14. Dynamics of iterative schemes for quadratic polynomial

    Science.gov (United States)

    Goyal, Komal; Prasad, Bhagwati

    2017-10-01

    Most of the fractals are generated by applying a map recursively to an initial point or a set of the space such as complex plane. An orbit of a point is a sequence of iteratively obtained points from it and the orbit is said to be diverging when its points grow unbounded. A set of all the points whose orbits are not diverging may be termed as a fractal. The fractal dynamics of the orbits depends on the rule of iteration also. In this paper, we study the dynamics of iterations for a number of iterative schemes. The results regarding the escape criteria for quadratic complex polynomials under various iteration procedures are established.

  15. Main chamber wall plasma loads in JET-ITER-like wall at high radiated fraction

    Directory of Open Access Journals (Sweden)

    C. Guillemaut

    2017-08-01

    Full Text Available Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW have reached a maximum radiative fraction (Frad of ∼75%. Divertor Langmuir probe (LP measurements in these discharges indicate, however, that less than ∼3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high Frad. In H-mode nitrogen-seeded discharges with Frad increasing from ∼30% to up to ∼75%, the main chamber wall particle fluence rises by a factor ∼3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at Frad ∼ 60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the fraction of the input power transported to the main chamber wall remains below ∼5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached conditions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem.

  16. RAMI analysis for ITER radial X-ray camera system

    Energy Technology Data Exchange (ETDEWEB)

    Qin, Shijun, E-mail: sjqin@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun; Chen, Kaiyun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Barnsley, Robin; Sirinelli, Antoine [ITER Organization, Route Vinon sur Verdon, CS 90046, 13067, St. Paul lez Durance, Cedex (France); Song, Yuntao; Lu, Kun; Yao, Damao; Chen, Yebin; Li, Shi; Cao, Hongrui; Yu, Hong; Sheng, Xiuli [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The functional analysis of the ITER RXC system was performed. • A failure modes, effects and criticality analysis of the ITER RXC system was performed. • The reliability and availability of the ITER RXC system and its main functions were calculated. • The ITER RAMI approach was applied to the ITER RXC system for technical risk control in the preliminary design phase. - Abstract: ITER is the first international experimental nuclear fusion device. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to mitigate all the possible failure of components in preparation for operation and maintenance. RAMI analysis of the ITER Radial X-ray Camera diagnostic (RXC) system during preliminary design phase was required, which insures the system with a very high performance to measure the X-ray emission and research the MHD of plasma with high accuracy on the ITER machine. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 3 main functions, 6 intermediate functions and 28 basic functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to calculate the reliability and availability of each function under assumption of operating conditions and failure data. Initial and expected scenarios were analyzed to define risk-mitigation actions. The initial availability of RXC system was 92.93%, while after optimization the expected availability was 95.23% over 11,520 h (approx. 16 months) which corresponds to ITER typical operation cycle. A Failure Modes, Effects and Criticality Analysis (FMECA) was performed to the system initial risk. Criticality charts highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. There are 28 risks for the initial state, including 8 major risks. No major risk remains after taking into

  17. Hybrid adsorptive membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  18. D and DR Reactors

    Data.gov (United States)

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

  19. Design and development of ITER high-frequency magnetic sensor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y., E-mail: Yunxing.Ma@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Fircroft Engineering, Lingley House, 120 Birchwood Point, Birchwood Boulevard, Warrington, WA3 7QH (United Kingdom); Vayakis, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Begrambekov, L.B. [National Research Nuclear University (MEPhI), 115409, Moscow, Kashirskoe shosse 31 (Russian Federation); Cooper, J.-J. [Culham Centre for Fusion Energy (CCFE), Abingdon, Oxfordshire OX14 3DB (United Kingdom); Duran, I. [IPP Prague, Za Slovankou 1782/3, 182 00 Prague 8 (Czech Republic); Hirsch, M.; Laqua, H.P. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Moreau, Ph. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Oosterbeek, J.W. [Eindhoven University of Technology (TU/e), PO Box 513, 5600 MB Eindhoven (Netherlands); Spuig, P. [CEA Cadarache, 13108 Saint Paul lez Durance Cedex (France); Stange, T. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Walsh, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-15

    Highlights: • ITER high-frequency magnetic sensor system has been designed. • Prototypes have been successfully manufactured. • Manufactured prototypes have been tested in various labs. • Test results experimentally validated the design. - Abstract: High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield.

  20. Combined Reactor and Microelectrode Measurements in Laboratory Grown Biofilms

    DEFF Research Database (Denmark)

    Larsen, Tove; Harremoës, Poul

    1994-01-01

    A combined biofilm reactor-/microelectrode experimental set-up has been constructed, allowing for simultaneous reactor mass balances and measurements of concentration profiles within the biofilm. The system consists of an annular biofilm reactor equipped with an oxygen microelectrode. Experiments...

  1. Reactor transient

    Energy Technology Data Exchange (ETDEWEB)

    Menegus, R.L.

    1956-05-31

    The authors are planning a calculation to be done on the Univac at the Louviers Building to estimate the effect of xenon transients, a high reactor power. This memorandum outlines the reasons why they prefer to do the work at Louviers rather than at another location, such as N.Y.U. They are to calculate the response of the reactor to a sudden change in position of the half rods. Qualitatively, the response will be a change in the rooftop ratio of the neutron flux. The rooftop ratio may oscillate with high damping, or, instead, it may oscillate for many cycles. It has not been possible for them to determine this response by hand calculation because of the complexity of the problem, and yet it is important for them to be certain that high power operation will not lead us to inherently unstable operation. Therefore they have resorted to machine computation. The system of differential equations that describes the response has seven dependent variables; therefore there are seven equations, each coupled with one or more of the others. The authors have discussed the problem with R.R. Haefner at the plant, and it is his opinion that the IBM 650 cannot adequately handle the system of seven equations because the characteristic time constants vary over a range of about 10{sup 8}. The Univac located at the Louviers Building is said to be satisfactory for this computation.

  2. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    Energy Technology Data Exchange (ETDEWEB)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  3. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  4. Experiment for search for sterile neutrino at SM-3 reactor

    Science.gov (United States)

    Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

    2016-11-01

    In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

  5. Reference worldwide model for antineutrinos from reactors

    Science.gov (United States)

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2015-03-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework, we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency. We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO +), and proposed (Juno, RENO-50, LENA, and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation, and detection processes are estimated using a Monte Carlo-based approach, which provides an overall site-dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes, and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of ten years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  6. On the Sequential Control of ITER Poloidal Field Converters for Reactive Power Reduction

    Science.gov (United States)

    Yuan, Hongwen; Fu, Peng; Gao, Ge; Huang, Liansheng; Song, Zhiquan; He, Shiying; Wu, Yanan; Dong, Lin; Wang, Min; Fang, Tongzhen

    2014-12-01

    Sequential control applied to the International Thermonuclear Experimental Reactor (ITER) poloidal field converter system for the purpose of reactive power reduction is the subject of this investigation. Due to the inherent characteristics of thyristor-based phase-controlled converter, the poloidal field converter system consumes a huge amount of reactive power from the grid, which subsequently results in a voltage drop at the 66 kV busbar if no measure is taken. The installation of a static var compensator rated for 750 MVar at the 66 kV busbar is an essential way to compensate reactive power to the grid, which is the most effective measure to solve the problem. However, sequential control of the multi-series converters provides an additional method to improve the natural power factor and thus alleviate the pressure of reactive power demand of the converter system without any additional cost. In the present paper, by comparing with the symmetrical control technique, the advantage of sequential control in reactive power consumption is highlighted. Simulation results based on SIMULINK are found in agreement with the theoretical analysis.

  7. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  8. Manufacture and test of prototype water pipe chase barrier in ITER Magnet Feeder system

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Kun, E-mail: lukun@ipp.ac.cn [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Wen, Xinjie; Liu, Chen; Song, Yuntao [Institute of Plasma Physics, Shushan Hu Road 350, Hefei, Anhui (China); Niu, Erwu [ITER China, 15B Fuxing Road, Beijing 100862 (China); Gung, Chenyu; Su, Man [ITER Organization, Route de Vinon-sur-Verdon – CS 90046, 13067 St Paul-lez-Durance Cedex (France)

    2016-11-01

    The Magnet Feeder system in the International Thermonuclear Experimental Reactor (ITER) deploys electrical currents and supercritical helium to the superconducting magnets and the magnet diagnostic signals to the operators. In the current design, the feeders located in the upper L3 level of the Tokamak gallery penetrate the Tokamak coolant water system vault, the biological shield and the cryostat. As a secondary confinement to contain the activated coolant water in the vault in the case of water pipe burst accident, a water barrier is welded between the penetration in the water pipe chase outer wall and the mid-plane of the vacuum jacket of the Feeder Coil Terminal Box (CTB). A thin-wall stainless steel diaphragm with an omega shape profile is welded around the CTB as the water barrier to endure 2 bar hydraulic pressure. In addition, the barrier is designed as a flexible compensator to withstand a maximum of 15 mm of axial displacement of the CTB in case of helium leak accident without failure. This paper presents the detail configuration, the manufacturing and assembly processes of the water barrier. Test results of the prototype water barrier under simulated accident conditions are also reported. Successful qualification of the design and manufacturing process of the water barrier lays a good foundation for the series production of this subsystem.

  9. Cryogenic instrumentation for ITER magnets

    Science.gov (United States)

    Poncet, J.-M.; Manzagol, J.; Attard, A.; André, J.; Bizel-Bizellot, L.; Bonnay, P.; Ercolani, E.; Luchier, N.; Girard, A.; Clayton, N.; Devred, A.; Huygen, S.; Journeaux, J.-Y.

    2017-02-01

    Accurate measurements of the helium flowrate and of the temperature of the ITER magnets is of fundamental importance to make sure that the magnets operate under well controlled and reliable conditions, and to allow suitable helium flow distribution in the magnets through the helium piping. Therefore, the temperature and flow rate measurements shall be reliable and accurate. In this paper, we present the thermometric chains as well as the venturi flow meters installed in the ITER magnets and their helium piping. The presented thermometric block design is based on the design developed by CERN for the LHC, which has been further optimized via thermal simulations carried out by CEA. The electronic part of the thermometric chain was entirely developed by the CEA and will be presented in detail: it is based on a lock-in measurement and small signal amplification, and also provides a web interface and software to an industrial PLC. This measuring device provides a reliable, accurate, electromagnetically immune, and fast (up to 100 Hz bandwidth) system for resistive temperature sensors between a few ohms to 100 kΩ. The flowmeters (venturi type) which make up part of the helium mass flow measurement chain have been completely designed, and manufacturing is on-going. The behaviour of the helium gas has been studied in detailed thanks to ANSYS CFX software in order to obtain the same differential pressure for all types of flowmeters. Measurement uncertainties have been estimated and the influence of input parameters has been studied. Mechanical calculations have been performed to guarantee the mechanical strength of the venturis required for pressure equipment operating in nuclear environment. In order to complete the helium mass flow measurement chain, different technologies of absolute and differential pressure sensors have been tested in an applied magnetic field to identify equipment compatible with the ITER environment.

  10. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    OpenAIRE

    José Tavares de Sousa; Keliana Dantas Santos; Israel Nunes Henrique; Danielle Patrício Brasil; Eclésio Cavalcante Santos

    2008-01-01

    The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was con...

  11. Nuclear analysis of ITER ports

    Energy Technology Data Exchange (ETDEWEB)

    Santoro, R.T.; Khripunov, V.; Lida, H.; Plenteda, D. [ITER Joint Central Team Garching, Muenchen (Germany); Sato, S. [Japan Atomic Energy Research Inst., Tokai-Mura (Japan); Inoue, T. [ITER Joint Central Team, naka (Japan)

    1998-07-01

    The steel/water blanket and vacuum vessel for the ITER Basic Performance Phase were designed reliable radiation shield for protecting the superconducting magnets, cryogenic systems and the cryostat from excessive heating from plasma neutrons and secondary gamma-rays. Proposed experiments, tests, diagnostics, and operation / maintenance equipment require numerous penetrations that weaken the effectiveness of the shield. However, comprehensive three-dimensional neutronic analyses and corresponding design developments, reported below, indicated that the key shielding requirements (nuclear heat deposition in the superconducting magnets {<=} 17 kW and residual equivalent dose rates {approx}100 {mu}Sv/h suitable for a personnel access) can be satisfied. (authors)

  12. Bioremediation of chlorpyrifos contaminated soil by two phase bioslurry reactor: Processes evaluation and optimization by Taguchi's design of experimental (DOE) methodology.

    Science.gov (United States)

    Pant, Apourv; Rai, J P N

    2018-04-15

    Two phase bioreactor was constructed, designed and developed to evaluate the chlorpyrifos remediation. Six biotic and abiotic factors (substrate-loading rate, slurry phase pH, slurry phase dissolved oxygen (DO), soil water ratio, temperature and soil micro flora load) were evaluated by design of experimental (DOE) methodology employing Taguchi's orthogonal array (OA). The selected six factors were considered at two levels L-8 array (2^7, 15 experiments) in the experimental design. The optimum operating conditions obtained from the methodology showed enhanced chlorpyrifos degradation from 283.86µg/g to 955.364µg/g by overall 70.34% of enhancement. In the present study, with the help of few well defined experimental parameters a mathematical model was constructed to understand the complex bioremediation process and optimize the approximate parameters upto great accuracy. Copyright © 2017 Elsevier Inc. All rights reserved.

  13. Nuclear reactor neutron shielding

    Science.gov (United States)

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

    2017-09-12

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

  14. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  15. IHadoop: Asynchronous iterations for MapReduce

    KAUST Repository

    Elnikety, Eslam Mohamed Ibrahim

    2011-11-01

    MapReduce is a distributed programming frame-work designed to ease the development of scalable data-intensive applications for large clusters of commodity machines. Most machine learning and data mining applications involve iterative computations over large datasets, such as the Web hyperlink structures and social network graphs. Yet, the MapReduce model does not efficiently support this important class of applications. The architecture of MapReduce, most critically its dataflow techniques and task scheduling, is completely unaware of the nature of iterative applications; tasks are scheduled according to a policy that optimizes the execution for a single iteration which wastes bandwidth, I/O, and CPU cycles when compared with an optimal execution for a consecutive set of iterations. This work presents iHadoop, a modified MapReduce model, and an associated implementation, optimized for iterative computations. The iHadoop model schedules iterations asynchronously. It connects the output of one iteration to the next, allowing both to process their data concurrently. iHadoop\\'s task scheduler exploits inter-iteration data locality by scheduling tasks that exhibit a producer/consumer relation on the same physical machine allowing a fast local data transfer. For those iterative applications that require satisfying certain criteria before termination, iHadoop runs the check concurrently during the execution of the subsequent iteration to further reduce the application\\'s latency. This paper also describes our implementation of the iHadoop model, and evaluates its performance against Hadoop, the widely used open source implementation of MapReduce. Experiments using different data analysis applications over real-world and synthetic datasets show that iHadoop performs better than Hadoop for iterative algorithms, reducing execution time of iterative applications by 25% on average. Furthermore, integrating iHadoop with HaLoop, a variant Hadoop implementation that caches

  16. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the

  17. Iterative Block Decision Feedback Equalization for MIMO Underwater Acoustic Communications

    Science.gov (United States)

    2010-06-01

    spread is excessively long; third, the Doppler spread effect is significantly large. Due to the harsh channel condition, demonstrated rates in UWA...orthogonal frequency division multiplexing ( OFDM ), are found in [2] and [3], [4], respectively. The time-domain (TD) receiver designs using channel...extensive experimental tests demonstrating the effectiveness of the iterative LE scheme. Hard-decision block decision-feedback equalizer (BDFE), which

  18. Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

    CERN Document Server

    Laitano, R F

    1987-01-01

    Design of the experimental apparatus to obtain a thermal neutron beam, intermediate-energy neutrons (2-144 keV) and high-energy photons (6 MeV) by means of the TRIGA reactor at the ENEA Casaccia center

  19. Measurement and calculations of long-lived radionuclide activity forming in the fast neutron field in some ITER construction steels

    Energy Technology Data Exchange (ETDEWEB)

    Pohorecki, W., E-mail: poho@agh.edu.pl [AGH University of Science and Technology, Faculty of Energy and Fuels, Al. Mickiewicza 30, 30-059 Krakow (Poland); Jodłowski, P. [AGH University of Science and Technology, Faculty of Physics and Applied Computer Science, Al. Mickiewicza 30, 30-059 Krakow (Poland); Pytel, K.; Prokopowicz, R. [National Centre for Nuclear Research, ul. Sołtana 7, 05-400 Otwock-Świerk (Poland)

    2014-10-15

    Highlights: • Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, in some ITER construction steels. • The neutron flux density was measured by means of activation foil method and unfolding technique. • Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TALYS-2011. • The activity measurements were done by means of gamma-ray spectrometry. - Abstract: Measurement and calculations of long-lived gamma-emitting radionuclide activity forming in the fission reactor fast neutron field were done, for some ITER construction steels. The activation was conducted in fast neutron irradiation channel of the MARIA research fission reactor (Poland). The dimensions of steel samples were 10 mm × 10 mm × 1 mm and mass was approximately 0.8 g. The neutron flux density was measured by means of activation foil method and unfolding technique; fraction of neutrons above 1 keV was 95%. The activation lasted 242 h and cooling took 100 days; the mean neutron flux density was 2.9E12 n/(cm{sup 2} s) (neutrons above 500 keV are 53% of total) whereas total fluency 2.53E18 cm{sup −2}. The activity measurements were done by means of gamma-ray spectrometry. Activity calculations were done by means of FISPACT-II code using the activation libraries EAF-2010 and TENDL-2011 and experimentally determined neutron flux. Measured activity of long-lived gamma emitting radionuclides was, in average, about 6.3 MBq/g 100 days after activation; the dominant radionuclides were {sup 58}Co and {sup 54}Mn (about 81% and 14% of total activity respectively). The C/E ratio differs for particular radionuclides and is in the range 0.86–0.92 for {sup 51}Cr, 0.93–1.21 for {sup 54}Mn, 0.77–0.98 for {sup 57}Co, 0.91–1.21 for {sup 58}Co, 1.17–1.27 for {sup 59}Fe, and 1.75–2.44 for {sup 60}Co.

  20. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  1. 3-D Monte Carlo analyses of the shielding system in a tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gallina, M.; Petrizzi, L.; Rado, V. (ENEA, Frascati (Italy). Centro Ricerche Energia)

    1990-01-01

    As part of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the shielding system performance in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) transport code (3-B version). The main issue is the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. ''Self generated weight windows'' and source biasing technique have been used to treat deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted. (author).

  2. 3-D Monte Carlo analyses of shielding system in tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gallina, M.; Petrizzi, L.; Rado, V.

    1990-09-01

    Within the framework of the ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations were carried out to assess system shielding performances in the basic machine configuration by means of the Monte Carlo Neutron Photon (MCNP) code (3-B version). The main issue concerns the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. 'Self generated weight windows' (w.w.) and source biasing techniques were used to treat the deep penetration through the bulk shield and streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted.

  3. Modelization of physical phenomena in research reactors with the help of new developments in transport methods, and methodology validation with experimental data; Modelisation des phenomenes physiques dans les reacteurs de recherche a l'aide de developpements realises dans les methodes de transport et qualification

    Energy Technology Data Exchange (ETDEWEB)

    Rauck, St

    2000-10-01

    The aim of this work is to develop a scheme for experimental reactors, based on transport equations. This type of reactors is characterized by a small core, a complex, very heterogeneous geometry and a large leakage. The possible insertion of neutron beams in the reflector and the presence of absorbers in the core increase the difficulty of the 3D-geometrical description and the physical modeling of the component parameters of the reactor. The Orphee reactor has been chosen for our study. Physical models (homogenization, collapsing cross section in few groups, albedo multigroup condition) have been developed in the APOLLO2 and CRONOS2 codes to calculate flux and power maps in a 3D-geometry, with different burnup and through transport equations. Comparisons with experimental measurements have shown the interest of taking into account anisotropy, steep flux gradients by using Sn methods, and on the other hand using a 12-group cross section library. The modeling of neutron beams has been done outside the core modeling through Monte Carlo calculations and with the total geometry, including a large thickness of heavy water. Thanks to this calculations, one can evaluate the neutron beams anti-reactivity and determinate the core cycle. We assure these methods more accurate than usual transport-diffusion calculations will be used for the conception of new research reactors. (author)

  4. The irace package: Iterated racing for automatic algorithm configuration

    Directory of Open Access Journals (Sweden)

    Manuel López-Ibáñez

    2016-01-01

    Full Text Available Modern optimization algorithms typically require the setting of a large number of parameters to optimize their performance. The immediate goal of automatic algorithm configuration is to find, automatically, the best parameter settings of an optimizer. Ultimately, automatic algorithm configuration has the potential to lead to new design paradigms for optimization software. The irace package is a software package that implements a number of automatic configuration procedures. In particular, it offers iterated racing procedures, which have been used successfully to automatically configure various state-of-the-art algorithms. The iterated racing procedures implemented in irace include the iterated F-race algorithm and several extensions and improvements over it. In this paper, we describe the rationale underlying the iterated racing procedures and introduce a number of recent extensions. Among these, we introduce a restart mechanism to avoid premature convergence, the use of truncated sampling distributions to handle correctly parameter bounds, and an elitist racing procedure for ensuring that the best configurations returned are also those evaluated in the highest number of training instances. We experimentally evaluate the most recent version of irace and demonstrate with a number of example applications the use and potential of irace, in particular, and automatic algorithm configuration, in general.

  5. CT radiation dose and iterative reconstruction techniques.

    Science.gov (United States)

    Padole, Atul; Ali Khawaja, Ranish Deedar; Kalra, Mannudeep K; Singh, Sarabjeet

    2015-04-01

    1. CT radiation dose optimization is one of the major concerns for the scientific community. 2. CT image quality is dependent on the selected image reconstruction algorithm. 3. Iterative reconstruction algorithms have reemerged with the potential of radiation dose optimization by lowering image noise. 4. Tube current is the most common parameter used to reduce radiation dose along with iterative reconstruction. 5. Tube potential (kV) is also used for dose optimization with iterative reconstruction in CT angiography protocols and small patients.

  6. Liquid metal blanket module testing and design for ITER/TIBER II

    Energy Technology Data Exchange (ETDEWEB)

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs.

  7. Tritium resources available for fusion reactors

    Science.gov (United States)

    Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

    2018-02-01

    The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

  8. LDPC-based iterative joint source-channel decoding for JPEG2000.

    Science.gov (United States)

    Pu, Lingling; Wu, Zhenyu; Bilgin, Ali; Marcellin, Michael W; Vasic, Bane

    2007-02-01

    A framework is proposed for iterative joint source-channel decoding of JPEG2000 codestreams. At the encoder, JPEG2000 is used to perform source coding with certain error-resilience (ER) modes, and LDPC codes are used to perform channel coding. During decoding, the source decoder uses the ER modes to identify corrupt sections of the codestream and provides this information to the channel decoder. Decoding is carried out jointly in an iterative fashion. Experimental results indicate that the proposed method requires fewer iterations and improves overall system performance.

  9. Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).

    Energy Technology Data Exchange (ETDEWEB)

    Parma, Edward J., Jr.

    2009-06-01

    The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.

  10. Overview on R&D and design activities for the ITER core charge exchange spectroscopy diagnostic system

    NARCIS (Netherlands)

    Biel, W.; Baross, T.; Bourauel, P.; Dunai, D.; Durkut, M.; Erdei, G.; Hawkes, N.; Hellermann, M. von; Hogenbirk, A.; Jaspers, R.; Kiss, G.; Klinkhamer, J.F.F.; Koning, J.F.; Kotov, V.; Krasikov, Y.; Krimmer, A.; Lischtschenko, O.; Litnovsky, A.; Marchuk, O.; Neubauer, O.; Offermanns, G.; Panin, A.; Patel, K.; Pokol, G.; Schrader, M.; Snijders, B.; Szabo, V.; Valk, N.J.C. van der; Voinchet, R.; Wolters, J.; Zoletnik, S.

    2011-01-01

    The ITER core charge exchange recombination spectroscopy (core CXRS) diagnostic system is designed to provide experimental access to various measurement quantities in the ITER core plasma such as ion densities, temperatures and velocities. The implementation of the approved CXRS diagnostic principle

  11. Managing Model Data Introduced Uncertainties in Simulator Predictions for Generation IV Systems via Optimum Experimental Design

    Energy Technology Data Exchange (ETDEWEB)

    Turinsky, Paul J [North Carolina State Univ., Raleigh, NC (United States); Abdel-Khalik, Hany S [North Carolina State Univ., Raleigh, NC (United States); Stover, Tracy E [North Carolina State Univ., Raleigh, NC (United States)

    2011-03-01

    An optimization technique has been developed to select optimized experimental design specifications to produce data specifically designed to be assimilated to optimize a given reactor concept. Data from the optimized experiment is assimilated to generate posteriori uncertainties on the reactor concept’s core attributes from which the design responses are computed. The reactor concept is then optimized with the new data to realize cost savings by reducing margin. The optimization problem iterates until an optimal experiment is found to maximize the savings. A new generation of innovative nuclear reactor designs, in particular fast neutron spectrum recycle reactors, are being considered for the application of closing the nuclear fuel cycle in the future. Safe and economical design of these reactors will require uncertainty reduction in basic nuclear data which are input to the reactor design. These data uncertainty propagate to design responses which in turn require the reactor designer to incorporate additional safety margin into the design, which often increases the cost of the reactor. Therefore basic nuclear data needs to be improved and this is accomplished through experimentation. Considering the high cost of nuclear experiments, it is desired to have an optimized experiment which will provide the data needed for uncertainty reduction such that a reactor design concept can meet its target accuracies or to allow savings to be realized by reducing the margin required due to uncertainty propagated from basic nuclear data. However, this optimization is coupled to the reactor design itself because with improved data the reactor concept can be re-optimized itself. It is thus desired to find the experiment that gives the best optimized reactor design. Methods are first established to model both the reactor concept and the experiment and to efficiently propagate the basic nuclear data uncertainty through these models to outputs. The representativity of the experiment

  12. COMPARISON OF HOLOGRAPHIC AND ITERATIVE METHODS FOR AMPLITUDE OBJECT RECONSTRUCTION

    Directory of Open Access Journals (Sweden)

    I. A. Shevkunov

    2015-01-01

    Full Text Available Experimental comparison of four methods for the wavefront reconstruction is presented. We considered two iterative and two holographic methods with different mathematical models and algorithms for recovery. The first two of these methods do not use a reference wave recording scheme that reduces requirements for stability of the installation. A major role in phase information reconstruction by such methods is played by a set of spatial intensity distributions, which are recorded as the recording matrix is being moved along the optical axis. The obtained data are used consistently for wavefront reconstruction using an iterative procedure. In the course of this procedure numerical distribution of the wavefront between the planes is performed. Thus, phase information of the wavefront is stored in every plane and calculated amplitude distributions are replaced for the measured ones in these planes. In the first of the compared methods, a two-dimensional Fresnel transform and iterative calculation in the object plane are used as a mathematical model. In the second approach, an angular spectrum method is used for numerical wavefront propagation, and the iterative calculation is carried out only between closely located planes of data registration. Two digital holography methods, based on the usage of the reference wave in the recording scheme and differing from each other by numerical reconstruction algorithm of digital holograms, are compared with the first two methods. The comparison proved that the iterative method based on 2D Fresnel transform gives results comparable with the result of common holographic method with the Fourier-filtering. It is shown that holographic method for reconstructing of the object complex amplitude in the process of the object amplitude reduction is the best among considered ones.

  13. Experience gathered in dismounting pressure vessel internals of the Kahl experimental reactor (VAK Kahl); Erfahrungen bei der Zerlegung der Druckgefaesseinbauten im VAK Kahl

    Energy Technology Data Exchange (ETDEWEB)

    Kalwa, H. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany); Eickelpasch, N. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany); Reiter, W. [Versuchsatomkraftwerk Kahl GmbH, Karlstein am Main (Germany)

    1995-11-01

    Mechanical cutting processes are preferred for dismantling of activated pressure vessel internals in the VAK Kahl. Up to now, sawing, nibbling, grinding and electroeroding have been tried. For a realistic under-water trial run on a scale of 1:1, an inactive experimental facility was erected in the empty turbine building of the VAK. (orig./DG) [Deutsch] Bei der Zerlegung der aktivierten RDB-Einbauten im Versuchsatomkraftwerk Kahl werden bevorzugt mechanische Trennverfahren angewendet. Erprobt und eingesetzt wurden bisher das Saegen, das Knabbern, das Schleifen und das Elektroerodieren. Fuer eine realitaetsnahe Erprobung unter Wasser im Massstab 1:1 wurde in VAK ein inaktiver Versuchsstand im bereits leergeraeumten Maschinenhaus erstellt. (orig./DG)

  14. Multiscale optical simulation settings: challenging applications handled with an iterative ray-tracing FDTD interface method.

    Science.gov (United States)

    Leiner, Claude; Nemitz, Wolfgang; Schweitzer, Susanne; Kuna, Ladislav; Wenzl, Franz P; Hartmann, Paul; Satzinger, Valentin; Sommer, Christian

    2016-03-20

    We show that with an appropriate combination of two optical simulation techniques-classical ray-tracing and the finite difference time domain method-an optical device containing multiple diffractive and refractive optical elements can be accurately simulated in an iterative simulation approach. We compare the simulation results with experimental measurements of the device to discuss the applicability and accuracy of our iterative simulation procedure.

  15. In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Damiano, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mehta, Chaitanya [Univ. of Tennessee, Knoxville, TN (United States); Collins, Price [Univ. of Tennessee, Knoxville, TN (United States); Lish, Matthew [Univ. of Tennessee, Knoxville, TN (United States); Cady, Brian [Univ. of Tennessee, Knoxville, TN (United States); Lollar, Victor [Univ. of Tennessee, Knoxville, TN (United States); de Wet, Dane [Univ. of Tennessee, Knoxville, TN (United States); Bayram, Duygu [Univ. of Tennessee, Knoxville, TN (United States)

    2015-12-15

    The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on

  16. On equivalence classes in iterative learning control

    NARCIS (Netherlands)

    Verwoerd, M.H.A.; Meinsma, Gjerrit; de Vries, Theodorus J.A.

    2003-01-01

    This paper advocates a new approach to study the relation between causal iterative learning control (ILC) and conventional feedback control. Central to this approach is the introduction of the set of admissible pairs (of operators) defined with respect to a family of iterations. Considered are two

  17. Iterative Brinkman penalization for remeshed vortex methods

    DEFF Research Database (Denmark)

    Hejlesen, Mads Mølholm; Koumoutsakos, Petros; Leonard, Anthony

    2015-01-01

    We introduce an iterative Brinkman penalization method for the enforcement of the no-slip boundary condition in remeshed vortex methods. In the proposed method, the Brinkman penalization is applied iteratively only in the neighborhood of the body. This allows for using significantly larger time...

  18. The ITER Thomson scattering core LIDAR diagnostic

    NARCIS (Netherlands)

    Naylor, G.A.; Scannell, R.; Beurskens, M.; Walsh, M.J.; Pastor, I.; Donné, A.J.H.; Snijders, B.; Biel, W.; Meszaros, B.; Giudicotti, L.; Pasqualotto, R.; Marot, L.

    2012-01-01

    The central electron temperature and density of the ITER plasma may be determined by Thomson scattering. A LIDAR topology is proposed in order to minimize the port access required of the ITER vacuum vessel. By using a LIDAR technique, a profile of the electron temperature and density can be

  19. Iterative methods for weighted least-squares

    Energy Technology Data Exchange (ETDEWEB)

    Bobrovnikova, E.Y.; Vavasis, S.A. [Cornell Univ., Ithaca, NY (United States)

    1996-12-31

    A weighted least-squares problem with a very ill-conditioned weight matrix arises in many applications. Because of round-off errors, the standard conjugate gradient method for solving this system does not give the correct answer even after n iterations. In this paper we propose an iterative algorithm based on a new type of reorthogonalization that converges to the solution.

  20. Controlled Iteration Grammars and Hyper-AFL's

    NARCIS (Netherlands)

    Asveld, P.R.J.

    1975-01-01

    We study $K$-iteration grammars equiped with a control on the application sequence of the substitutions, called $(\\Gamma,K)$-iteration grammars. In fact, the control is a language over the indices of the substitutions, prescribing the specific order in which one has to apply the different

  1. ITER Fast Ion Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Bindslev, Henrik; Meo, Fernando; Korsholm, Søren Bang

    for measurements of the confined fusion alpha particles in ITER set by the ITER team. Then we outline the considerations, which enter into the selection and evaluation of CTS systems. System definition includes choice of probe frequency, geometry of probe and receiver beam patterns and probe power, but ultimately...

  2. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  3. Techniques in Iterative Proton CT Image Reconstruction

    CERN Document Server

    Penfold, Scott

    2015-01-01

    This is a review paper on some of the physics, modeling, and iterative algorithms in proton computed tomography (pCT) image reconstruction. The primary challenge in pCT image reconstruction lies in the degraded spatial resolution resulting from multiple Coulomb scattering within the imaged object. Analytical models such as the most likely path (MLP) have been proposed to predict the scattered trajectory from measurements of individual proton location and direction before and after the object. Iterative algorithms provide a flexible tool with which to incorporate these models into image reconstruction. The modeling leads to a large and sparse linear system of equations that can efficiently be solved by projection methods-based iterative algorithms. Such algorithms perform projections of the iterates onto the hyperlanes that are represented by the linear equations of the system. They perform these projections in possibly various algorithmic structures, such as block-iterative projections (BIP), string-averaging...

  4. RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part II: Analysis of ITER plasma facing components

    Science.gov (United States)

    Federici, Gianfranco; Raffray, A. René

    1997-04-01

    The transient thermal model RACLETTE (acronym of Rate Analysis Code for pLasma Energy Transfer Transient Evaluation) described in part I of this paper is applied here to analyse the heat transfer and erosion effects of various slow (100 ms-10 s) high power energy transients on the actively cooled plasma facing components (PFCs) of the International Thermonuclear Experimental Reactor (ITER). These have a strong bearing on the PFC design and need careful analysis. The relevant parameters affecting the heat transfer during the plasma excursions are established. The temperature variation with time and space is evaluated together with the extent of vaporisation and melting (the latter only for metals) for the different candidate armour materials considered for the design (i.e., Be for the primary first wall, Be and CFCs for the limiter, Be, W, and CFCs for the divertor plates) and including for certain cases low-density vapour shielding effects. The critical heat flux, the change of the coolant parameters and the possible severe degradation of the coolant heat removal capability that could result under certain conditions during these transients, for example for the limiter, are also evaluated. Based on the results, the design implications on the heat removal performance and erosion damage of the variuos ITER PFCs are critically discussed and some recommendations are made for the selection of the most adequate protection materials and optimum armour thickness.

  5. An iterative method for selecting degenerate multiplex PCR primers.

    Science.gov (United States)

    Souvenir, Richard; Buhler, Jeremy; Stormo, Gary; Zhang, Weixiong

    2007-01-01

    Single-nucleotide polymorphism (SNP) genotyping is an important molecular genetics process, which can produce results that will be useful in the medical field. Because of inherent complexities in DNA manipulation and analysis, many different methods have been proposed for a standard assay. One of the proposed techniques for performing SNP genotyping requires amplifying regions of DNA surrounding a large number of SNP loci. To automate a portion of this particular method, it is necessary to select a set of primers for the experiment. Selecting these primers can be formulated as the Multiple Degenerate Primer Design (MDPD) problem. The Multiple, Iterative Primer Selector (MIPS) is an iterative beam-search algorithm for MDPD. Theoretical and experimental analyses show that this algorithm performs well compared with the limits of degenerate primer design. Furthermore, MIPS outperforms an existing algorithm that was designed for a related degenerate primer selection problem.

  6. A new automatic baseline correction method based on iterative method

    Science.gov (United States)

    Bao, Qingjia; Feng, Jiwen; Chen, Fang; Mao, Wenping; Liu, Zao; Liu, Kewen; Liu, Chaoyang

    2012-05-01

    A new automatic baseline correction method for Nuclear Magnetic Resonance (NMR) spectra is presented. It is based on an improved baseline recognition method and a new iterative baseline modeling method. The presented baseline recognition method takes advantages of three baseline recognition algorithms in order to recognize all signals in spectra. While in the iterative baseline modeling method, besides the well-recognized baseline points in signal-free regions, the 'quasi-baseline points' in the signal-crowded regions are also identified and then utilized to improve robustness by preventing the negative regions. The experimental results on both simulated data and real metabolomics spectra with over-crowded peaks show the efficiency of this automatic method.

  7. Hydrogen Spectral Line Shape Formation in the SOL of Fusion Reactor Plasmas

    Directory of Open Access Journals (Sweden)

    Valery S. Lisitsa

    2014-05-01

    Full Text Available The problems related to the spectral line-shape formation in the scrape of layer (SOL in fusion reactor plasma for typical observation chords are considered. The SOL plasma is characterized by the relatively low electron density (1012–1013 cm−3 and high temperature (from 10 eV up to 1 keV. The main effects responsible for the line-shape formation in the SOL are Doppler and Zeeman effects. The main problem is a correct modeling of the neutral atom velocity distribution function (VDF. The VDF is determined by a number of atomic processes, namely: molecular dissociation, ionization and charge exchange of neutral atoms on plasma ions, electron excitation accompanied by the charge exchange from atomic excited states, and atom reflection from the wall. All the processes take place step by step during atom motion from the wall to the plasma core. In practice, the largest contribution to the neutral atom radiation emission comes from a thin layer near the wall with typical size 10–20 cm, which is small as compared with the minor radius of modern devices including international test experimental reactor ITER (radius 2 m. The important problem is a strongly non-uniform distribution of plasma parameters (electron and ion densities and temperatures. The distributions vary for different observation chords and ITER operation regimes. In the present report, most attention is paid to the problem of the VDF calculations. The most correct method for solving the problem is an application of the Monte Carlo method for atom motion near the wall. However, the method is sometimes too complicated to be combined with other numerical codes for plasma modeling for various regimes of fusion reactor operation. Thus, it is important to develop simpler methods for neutral atom VDF in space coordinates and velocities. The efficiency of such methods has to be tested via a comparison with the Monte Carlo codes for particular plasma conditions. Here a new simplified method

  8. Continuous esterification to produce biodiesel by SPES/PES/NWF composite catalytic membrane in flow-through membrane reactor: experimental and kinetic studies.

    Science.gov (United States)