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Sample records for braunschweig experimental reactor

  1. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

  2. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  3. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  4. Irrigation of treated wastewater in Braunschweig, Germany

    DEFF Research Database (Denmark)

    Ternes, T.A.; Bonerz, M.; Herrmann, N.;

    2007-01-01

    of digested sludge, because many polar compounds do not sorb to sludge and lipophilic compounds are not mobile in the soil-aquifer. Most of the selected PPCPs were never detected in any of the lysimeter or groundwater samples, although they were present in the treated wastewater irrigated onto the fields...... of the sewage treatment plant (STP) of Braunschweig is used for irrigation, while during summer digested sludge is mixed with the effluent. In the present case study six wells and four lysimeters located in one of the irrigated agricultural fields were monitored with regard to the occurrence of 52....... In the groundwater and lysimeter samples primarily the ICM diatrizoate and iopamidol, the antiepileptic carbamazepine and the antibiotic sulfamethoxazole were detected up to several mu g l(-1), while the acidic pharmaceuticals, musk fragrances, estrogens and betablockers were likely sorbed or transformed while...

  5. 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.

  6. Future experimental programmes in the CROCUS reactor

    OpenAIRE

    Lamirand, Vincent Pierre; Hursin, Mathieu; Perret, Grégory; Frajtag, Pavel; Pakari, Oskari; Pautz, Andreas

    2016-01-01

    CROCUS is a teaching and research zero-power reactor operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology (EPFL). Three new experimental programmes are scheduled for the forthcoming years. The first programme consists in an experimental investigation of mechanical noise induced by fuel rods vibrations. An in-core device has been designed for allowing the displacement of up to 18 uranium metal fuel rods in the core periphery. ...

  7. 9 CFR 319.182 - Braunschweiger and liver sausage or liverwurst.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Braunschweiger and liver sausage or... Sausage § 319.182 Braunschweiger and liver sausage or liverwurst. (a) “Braunschweiger” is a cooked sausage... the following: “Braunschweiger—A Liver Sausage,” “Braunschweiger—A Liverwurst,” or...

  8. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  9. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in a previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes were considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the ''in-situ'' replacement of first walls using atomic coating processes were considered. The vapor deposition of carbon was shown to be promising

  10. Fusion experimental power reactor (EPR) design tasks

    International Nuclear Information System (INIS)

    Several key physics and technology problem areas which were identified in the previous Experimental Power Reactor study were investigated. These were plasma confinement, plasma heating, reactor refueling, and reactor first wall regeneration. The plasma confinement experimental studies showed no instabilities or enhanced transport in the trapped ion regime. The RF heating experiments indicated that RF could produce highly efficient plasma heating. Two reactor refueling schemes was considered in a theoretical analysis: the first was the convective transport from the cold plasma blanket to the plasma interior and the second was the use of high speed frozen pellets to carry the fuel to the plasma interior. Both schemes were shown to be feasible. Finally, the in-situ replacement of first walls using atomic coating processes was considered. The vapor deposition of carbon was shown to be promising

  11. Overview of FER (Fusion Experimental Reactor)

    International Nuclear Information System (INIS)

    The FER (Fusion Experimental Reactor) project is proposed to construct a next generation tokamak machine in Japan, in order to take a leadership in realizing a fusion reactor under international cooperation. The FER is the machine, which comes in between the present large tokamak machines like the JT-60 and the DEMO reactor for power generation. The mission of the FER is to realize a long controlled burn and to develop and test major fusion component technologies, super conducting magnet and breeding blanket and so on, that is, to demonstrate the engineering feasibility of a fusion reactor. The conceptual design of the FER was started in 1980. In April 1988, a new organization (Fusion Experimental Reactor Team) was constructed to support the ITER activities and also to design the FER. In order to make the FER and the ITER complementary, the FER concept was reconsidered. The FER described in this report is a new version, and the conceptual design will be finished in December, 1990. (author)

  12. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  13. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

  14. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

    The invention claims a means for the adjustment of control rods in experimental reactors with a continuously variable pitch of the fuel element spacer. The proposed device permits obtaining maximum variability in the physical modelling of nuclear power reactor cores in experimental reactors. (F.M.)

  15. Das DFG-Sondersammelgebiet Pharmazie der Universitätsbibliothek Braunschweig / DFG-Special Subject Collection Pharmacy of Braunschweig University Library

    Directory of Open Access Journals (Sweden)

    Wulle, Stefan

    2011-09-01

    Full Text Available The Special Subject Collection of Braunschweig University Library is the most extensive collection of pharmaceutical literature in Central Europe. For some 62 years the Deutsche Forschungsgemeinschaft, DFG, has supported the Special Subject Collection of Pharmacy that now involves about 51,000 monographs and more than 600 journal subscriptions. The aim of the Virtual Library of Pharmacy is the global detection of conventional and electronic sources, internet items and database contents for the subject of pharmacy through a single portal.

  16. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  17. Second preliminary design of JAERI experimental fusion reactor (JXFR)

    International Nuclear Information System (INIS)

    Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

  18. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  19. Instrumentation and control improvements at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  20. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Jianzhong Zhao; Yaqin Tian; Yangsheng Zhao; Wenping Cheng

    2015-01-01

    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 ...

  3. 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).

  4. A method of safety assurance for fusion experimental reactor

    International Nuclear Information System (INIS)

    The present report describes safety assurance method for fusion experimental reactor. The ALARA (As Low As Reasonably Achievable) principle for a normal condition and the defence in depth principle for states deviated from the normal condition can be used as basic principles of safety assurance of the reactor. The method includes safety design for systems, importance categorization method to impose suitable demands to their systems, safety evaluation method to validate the design and application of the method. It is considered that this method can be a strong candidate for safety assurance method. (author)

  5. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of conceptual design study on plant systems for the Fusion Experimental Reactor (FY86 FER). Design studies for FER plant systems have been continued from FY85, especially for design modifications made in accordance with revisions of plasma scaling parameters and system improvements. This report describes 1) system construction, 2) site and reactor building plan, 3) repaire and maintenance system, 4) tritium circulation system, 5) heating, ventilation and air conditioning system, 6) tritium clean-up system, 7) cooling and baking system, 8) waste treatment and storage system, 9) control system, 10) electric power system, 11) site factory plan, all of which are a part of FY86 design work. The plant systems described in this report generally have been based on the FY86 FER (ACS Reactor) which is an one of the six candidates for FER. (author)

  6. Development and Application of Agricultural Composting Reactor Experimental Apparatus

    Directory of Open Access Journals (Sweden)

    Jizong Jiao

    2015-06-01

    Full Text Available In this study, we have a research of the development and application of agricultural composting reactor experimental apparatus. In agriculture, conversion of straw, sludge and other junk to agricultural fertilizer have a broad market prospect, it not only benefit solid waste recycling but also increase food production and food quality. Currently, the existing composting reactor had several shortcomings in practical application process, such as uneven mixing, long composting reaction cycle, producing odor and low maturity. Considering of materials and spatial structure, ventilation system, stirring system, we designed and developed a new experimental device aerobic composting reactor with relatively simple structure. We used the material in experimental device to doing a series of experiments, it include stir even, temperature changes, moisture changes, changes of organic matter, ammonia nitrogen change, change of nitrate nitrogen, PH values and so on. Experimental results showed that standard organic fertilizer could be produced by the device. Moreover, it had more complete degradation of organic matter to improve the quality of the compost product. Experiments also showed that compared with other existing devices, using the device could ferment evenly, increase the speed of biochemical reactions and reduce the fermentation time.

  7. Safety review, assessment and inspection on research reactors, experimental reactors and nuclear heating reactors

    International Nuclear Information System (INIS)

    The NNSA and its regional office step further strengthened the regulation on the safety of in-service research reactors in 1996. A lot of work has been done on the supervision of safe in rectifying the review and assessment of modified items, the review of operational documents, the treatment of accidents, the establishment of the system for operational experience feedback, daily and routine inspection on nuclear safety. The internal management of the operating organization on nuclear safety was further strengthened, nuclear safety culture was further enhanced, the promotion in nuclear safety and the safety situation for in-service research reactors were improved

  8. WWER reactor fuel performance, modelling and experimental support. Proceedings

    International Nuclear Information System (INIS)

    This publication is a compilation of 36 papers presented at the International Seminar on WWER Reactor Fuel Performance, Modelling and Experimental Support, organised by the Institute for Nuclear Research and Nuclear Energy (BG), in cooperation with the International Atomic Energy Agency. The Seminar was attended by 76 participants from 16 countries, including representatives of all major Russian plants and institutions responsible for WWER reactor fuel manufacturing, design and research. The reports are grouped in four chapters: 1) WWER Fuel Performance and Economics: Status and Improvement Prospects: 2) WWER Fuel Behaviour Modelling and Experimental Support; 3) Licensing of WWER Fuel and Fuel Analysis Codes; 4) Spent Fuel of WWER Plants. The reports from the corresponding four panel discussion sessions are also included. All individual papers are recorded in INIS as separate items

  9. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  10. A neutronradiography facility based on an experimental reactor

    OpenAIRE

    THOMAS DIMITRIOS; J. G. Fantidis; NICOLAOU G.

    2014-01-01

    A thermal Neutron Radiography (NR) facility based on the use of thermal neutron flux, generated by the PULSTAR experimental reactor, has been designed and simulated using the MCNPX code. The key objective of the proposed facility is to deliver thermal neutron flux in this range for variable values of L/D ratio, instantaneously with acceptable values for all NR parameters. Thus, with suitable aperture and collimators designs, optimization for the parameters for thermal NR was achieved, for a w...

  11. Parameter estimation and optimal experimental design in flow reactors

    OpenAIRE

    Carraro, Thomas

    2005-01-01

    In this work we present numerical techniques, based on the finite element method, for the simulation of reactive flows in a chemical flow reactor as well as for the identification of the kinetic of the reactions using measurements of observable quantities. We present the case of a real experiment in which the reaction rate is estimated by means of concentration measurements. We introduce methods for the optimal experimental design of experiments in the context of reactive flows modeled by par...

  12. Projections for a steady-state tokamak reactor based on the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Devoto, R.S.; Barr, W.L.; Bulmer, R.H.; Campbell, R.B.; Fenstermacher, M.E.; Lee, J.D.; Logan, B.G.; Miller, J.R.; Reginato, L.L. (Lawrence Livermore National Lab., Livermore, CA (US)); Krakowski, R.A.; Miller, R.L. (Los Alamos National Lab., Los Alamos, NM (US))

    1991-03-01

    This paper examines the extensions of the physics and engineering guidelines for the International Thermonuclear Experimental Reactor (ITER) device needed for acceptable operating points for a steady-state tokamak power reactor. Noninductive current drive is provided in steady state by high-energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma, and bootstrap current. Three different levels of extension of the ITER physics/engineering guidelines, with differing assumptions on the possible plasma beta, elongation, and aspect ratio, are considered for power reactor applications. Plasma gain Q = fusion power/input power in excess of 20 and average neutron wall fluxes from 2.3 to 3.6 MW/m{sup 2} are predicted in devices with major radii varying from 7.0 to 6.0 m and aspect ratios from 2.9 to 4.3.

  13. A Study of Reactor Neutrino Monitoring at Experimental Fast Reactor JOYO

    CERN Document Server

    Furuta, H; Hara, T; Haruna, T; Ishihara, N; Ishitsuka, M; Ito, C; Katsumata, M; Kawasaki, T; Konno, T; Kuze, M; Maeda, J; Matsubara, T; Miyata, H; Nagasaka, Y; Nitta, K; Sakamoto, Y; Suekane, F; Sumiyoshi, T; Tabata, H; Takamatsu, M; Tamura, N

    2011-01-01

    We carried out a study of neutrino detection at the experimental fast reactor JOYO using a 0.76 tons gadolinium loaded liquid scintillator detector. The detector was set up on the ground level at 24.3m from the JOYO reactor core of 140MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11\\pm1.24(stat.)\\pm0.46(syst.)events/day. Although the statistical significance of the measurement was not enough, the background in such a compact detector at the ground level was studied in detail and MC simulation was found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  14. 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.

  15. TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

  16. Conceptual design study of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M2. 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

  17. Experimental Study of a Photocatalytic Reactor for Trace Formaldehyde Removal

    Institute of Scientific and Technical Information of China (English)

    LIU Hong-min; LIAN Zhi-wei; YE Xiao-jiang; SHANG-GUAN Wen-feng

    2005-01-01

    Formaldehyde is the key contaminant influencing building occupants' health in indoor environment. In order to reduce occupants' exposures to formaldehyde, a newly designed photocatalytic reactor was applied in a dynamic HVAC(heating, ventilation and air conditioning) system. The experiments were carried out for the removal of formaldehyde present in air at low parts per million (ppm) concentrations.The initial formaldehyde concentrations were set as1.59 ppm and 0.27 ppm respectively, based on the formaldehyde levels in the polluted places. Experimental results show that the photocatalytic reactor is effective on formaldehyde photodegradation, causes a low pressure drop, and does not make the second pollution of ozone. The kinetic analysis indicates that the kinetics for oxidation processes can be fitted well by a pseudo-first-order kinetic model deduced from Langmuir - Hinshelwood (L-H) model.

  18. Experimental studies on catalytic hydrogen recombiners for light water reactors

    International Nuclear Information System (INIS)

    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.)

  19. Shutdown and Closure of the Experimental Breeder Reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor - II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated lay-up plan defining the system end state, as well as instructions for achieving the lay-up condition. A goal of system-by-system lay-up is to minimize surveillance and

  20. Shutdown and closure of the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m3 (86,000 gallons) of sodium and the secondary system contained 50 m3 (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance and

  1. Radiochemical characterization of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Nuclear reactors which have in-built graphite may receive a high neutron dose for a long period. Depending on the chemical composition of the graphite, numerous activation products may result. In addition, the amount of fission product contamination will depend on the location of the graphite. The migration of fission products may be supported by the high temperatures which occur in high-temperature reactors. At the Juelich 15 MWe high-temperature gas-cooled experimental AVR (Arbeitsgemeinschaft Versuchsreaktor) reactor, two different types of nuclear graphite had been in use. High-purity graphite was used as a basic material for core structures of the AVR. Insulation layers of carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For various reasons it is important to know the degree of contamination of graphite and carbon bricks from activation and fission products. The optimum method for nuclear graphite analysis in decommissioning is by incineration. Volatile activities (14C, 3H, 36Cl, ...) have to be captured for analysis. In cases where dust-like samples are handled, the incineration furnace has to be small enough to be operated in a glove-box. The resulting ashes can be used for determining all non-volatile nuclides by different radiochemical methods. In early 1999, some graphite and carbon brick samples from the AVR reactor were obtained by drilling. The samples were then analysed in the laboratories at the Juelich research centre. For incineration a vertical quartz tube was used which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl were captured in washing bottles. After further preparation, they were analysed by liquid scintillation counting (LSC). After dissolving the ashes, the elements were separated by ion exchange, extraction methods and HPLC. The radionuclides were then determined by alpha-spectrometry, LSC, low

  2. Non destructive examination of Reactor DR-3. Reactor wall, horisontal experimental tubes, up- and down comers

    International Nuclear Information System (INIS)

    The initial scope of work was to perform thickness/corrosion measurements of one up-comer and one down-comer, perform thickness/corrosion measurements in selected areas of the reactor wall and horizontal experimental pipes inside the reactor. Furthermore the lower circumferential weld and the connected longitudinal weld should be inspected to the extent possible, without major changes of the manipulator. Eddy current was performed in the same areas. Also hardness tests were carried out in four positions inside the reactor. Due to the outcome of the above examinations, additional metallurgical and dye penetrant examinations (PT) were carried out. The examination of the up- and down comers showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The examination of the horizontal tubes showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The hardness test showed increased hardness compared to calibration values. The examination of the reactor wall base material revealed several indications located in different depths in the plate. Some indications have been proved to be connected to the inner surface, while most indications appear to be either inclusions or areas corroded from the outside reactor wall. Minimum measured wall thickness is between 4.2 and 11.0 mm. There is, however, no evidence that these values are caused by corrosion at the outer reactor surface. The ET showed no signs of service induced cracks. The hardness test showed values close to calibration values. The extensive number of indications has resulted in additional

  3. Experimental and theoretical investigation of anaerobic fluidized bed biofilm reactors

    Directory of Open Access Journals (Sweden)

    M. Fuentes

    2009-09-01

    Full Text Available This work presents an experimental and theoretical investigation of anaerobic fluidized bed reactors (AFBRs. The bioreactors are modeled as dynamic three-phase systems. Biochemical transformations are assumed to occur only in the fluidized bed zone. The biofilm process model is coupled to the system hydrodynamic model through the biofilm detachment rate; which is assumed to be a first-order function of the energy dissipation parameter and a second order function of biofilm thickness. Non-active biomass is considered to be particulate material subject to hydrolysis. The model includes the anaerobic conversion for complex substrate degradation and kinetic parameters selected from the literature. The experimental set-up consisted of two mesophilic (36±1ºC lab-scale AFBRs (R1 and R2 loaded with sand as inert support for biofilm development. The reactor start-up policy was based on gradual increments in the organic loading rate (OLR, over a four month period. Step-type disturbances were applied on the inlet (glucose and acetic acid substrate concentration (chemical oxygen demand (COD from 0.85 to 2.66 g L-1 and on the feed flow rate (from 3.2 up to 6.0 L d-1 considering the maximum efficiency as the reactor loading rate switching. The predicted and measured responses of the total and soluble COD, volatile fatty acid (VFA concentrations, biogas production rate and pH were investigated. Regarding hydrodynamic and fluidization aspects, variations of the bed expansion due to disturbances in the inlet flow rate and the biofilm growth were measured. As rate coefficients for the biofilm detachment model, empirical values of 3.73⋅10(4 and 0.75⋅10(4 s² kg-1 m-1 for R1 and R2, respectively, were estimated.

  4. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  5. Experimental studies of fission properties utilized in reactor design

    International Nuclear Information System (INIS)

    Experimental studies of fission properties utilized in reactor design. A programme of experimental studies of fission parameters useful in reactor design is described including the following: (a) The periods and yields of delayed-neutron groups emitted following the neutron-induced fission of Pu241 are measured. Evidence for systematic isotopic dependence of delayed-neutron yields is presented. An experimental investigation of the relation between the time behaviour of delayed-neutron emission and the energy of the incident neutron inducing fission is described. (b) The cross-section for the inducing, of fission in Am243, Pu242 and Pu241 with neutrons in the energy range 0.030 to 1.8 MeV is measured. Emphasis is placed upon the detailed dependence of the fission cross-section on the incident-neutron energy. The absolute values of the cross-sections are given to a precision of ∼25%. (c) Detailed results of a measurement of the Pu241 fission-neutron spectrum are given, including the spectral shape and average fission-neutron energy. Techniques and methods of measuring prompt-fission-neutron spectra are described. (d) The dependence of #-v# (the average number of neutrons emitted per fission) of U235 on the incident neutron energy is measured from 100 keV to 1.6 MeV. #-v# of U238 and other fissile isotopes is compared to #-v# of U235 (thermal). The relative precision of the measurements is #>approx#1.2%. (author)

  6. Detection efficiency of the neutron detector BELEN-48 measured at the PTB Braunschweig

    Energy Technology Data Exchange (ETDEWEB)

    Marta, Michele [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); Agramunt, Jorge; Tain, Jose Luis [IFIC-CSIC University of Valencia, Valencia (Spain); Caballero-Folch, Roger; Cortes, Guillem; Riego, Albert [INTE-DFEN, Universitat Politecnica de Catalunya, Barcelona (Spain); Dillmann, Iris [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); II. Physikalisches Institut, Justus-Liebig Universitaet Giessen (Germany); TRIUMF, Vancouver (Canada); Erhard, Martin; Giesen, Ulrich; Nolte, Ralf; Roettger, Stefan [Physikalisch-Technische Bundesanstalt (PTB), Braunschweig (Germany); Fraile, Luis M. [Universidad Complutense de Madrid (Spain)

    2014-07-01

    The BEta-deLayEd Neutron detector BELEN-48 is a highly efficient detector of β-delayed neutrons, for nuclear structure, nuclear astrophysics and reactor studies. It consists of 48 {sup 3}He proportional counters arranged in a polyethylene matrix in a way that the detection efficiency remains constant for neutron energies from thermal up to a few MeV. In order to validate MCNPX simulations, the detection efficiency has been calibrated with well-known (p,n) and (α,n) reactions on {sup 7}Li, {sup 13}C and {sup 51}V producing neutrons with energies between 0.1 and 5 MeV. The experiment has been performed at the neutron metrology facility of PTB, which allowed the measurement of yields and angular distributions with a calibrated monitor. The new results indicate anisotropies, which are not reported in literature and have been taken into account to obtain the experimental efficiencies for BELEN.

  7. Oak Ridge Tokamak experimental power reactor study scoping report

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.

    1977-03-01

    This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR.

  8. A neutronradiography facility based on an experimental reactor

    Directory of Open Access Journals (Sweden)

    D. T. Thomas

    2015-06-01

    Full Text Available A thermal Neutron Radiography (NR facility based on the use of thermal neutron flux, generated by the PULSTAR experimental reactor, has been designed and simulated using the MCNPX code. The key objective of the proposed facility is to deliver thermal neutron flux in this range for variable values of L/D ratio, instantaneously with acceptable values for all NR parameters. Thus, with suitable aperture and collimators designs, optimization for the parameters for thermal NR was achieved, for a wide range of the collimator ratio. The short time requirements for obtaining the radiography images justify the use of the proposed system for ‘real time radiography’. The system was designed under the limitation that the total Dose Equivalent Rate does not exceed at the external shield surface the limit recommended by ICRP-26.

  9. Experimental Investigation of Gas-Lift Use in Nuclear Reactors

    International Nuclear Information System (INIS)

    This work briefly describes the selection of type of a two-phase flow, suitable for intensifying of a natural flow of nuclear reactors with liquid fuel – cooling mixture molten salts and the description of a „Two-phase flow demonstrator“ (TFD) used for experimental study of the „gas-lift“ system, and its influence on the support of natural convection. The experimental device works with water and the air is used as a gas. The used perspex limits the temperature to 60°C. There are stated relations for the description of a natural flow in model device and relations for determination of suitable liquid/gas ratio of the gas-lift in the study. There is described the measuring device and the application of the TFD sensor. The flow rate of water is measured by the induction flow meter that gives a voltage signal, which is brought into a computer for processing. Measuring of the velocity distribution and the size of the bubbles is performed by using the PIV method (Particle Image Velocimetry). There was created a model of dispersive bubble flow for application in nuclear reactors. The basic calculation is performed by using the homogeneous flow, where is considered, that the velocity of the fluid and the gas is equal and there is measured the relative share of the gas in homogeneous mixture with the fluid for this case. There are considered the temperature, pressure and flow rate velocity changes of the fluid and gas in the gas-lift cylinder and their influence on the size and velocity of the bubbles for the heat and mass transport of this mixture by the gas-lift cylinder. (author)

  10. Experimental assessment of computer codes used for safety analysis of integral reactors

    Energy Technology Data Exchange (ETDEWEB)

    Falkov, A.A.; Kuul, V.S.; Samoilov, O.B. [OKB Mechanical Engineering, Nizhny Novgorod (Russian Federation)

    1995-09-01

    Peculiarities of integral reactor thermohydraulics in accidents are associated with presence of noncondensable gas in built-in pressurizer, absence of pumped ECCS, use of guard vessel for LOCAs localisation and passive RHRS through in-reactor HX`s. These features defined the main trends in experimental investigations and verification efforts for computer codes applied. The paper reviews briefly the performed experimental investigation of thermohydraulics of AST-500, VPBER600-type integral reactors. The characteristic of UROVEN/MB-3 code for LOCAs analysis in integral reactors and results of its verification are given. The assessment of RELAP5/mod3 applicability for accident analysis in integral reactor is presented.

  11. Experimental study of neutrino oscillations at a fission reactor

    International Nuclear Information System (INIS)

    The energy spectrum of neutrinos from a fission reactor was studied with the aim of gaining information on neutrino oscillations. The well shielded detector was set up at a fixed position of 8.76 m from the point-like core of the Laue-Langevin reactor in an antineutrino flux of 9.8 x 1011cm-2s-1. The target protons in the reaction antiνsub(e)p → e+n were provided by liquid scintillation counters (total volume of 377l) which also served as positron detectors. The product neutrons moderated in the scintillator were detected by 3He wire chambers. A coincidence signature was required between the prompt positron and the delayed neutron events. The positron energy resolution was 18% FWHM at 0.91 MeV. The signal-to-background ratio was better than one to one between 2 MeV and 6 MeV positron energy. At a counting rate of 1.58 counts per hour, 4890+-180 neutrino induced events were detected. The shape of the measured positron spectrum was analyzed in terms of the parameters Δ2 and sin2 2theta for two-neutrino oscillations. The experimental data are consistent with no oscillations. An upper limit of 0.15 eV2 (90% c.l.) for the mass-squared differences Δ2 of the neutrinos was obtained, assuming maximum mixing of the two neutrino states. The ratio of the measured to the expected integral yield of positrons assuming no oscillations was determined to be ∫Ysub(exp)/∫Ysub(th) = 0.955+-0.035 (statistical)+-0.110 (systematic)

  12. 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.

  13. Experimental study of coal topping process in a downer reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.G.; Lu, X.S.; Yao, J.X.; Lin, W.G.; Cui, L.J. [Chinese Academy of Science, Beijing (China). Inst. of Processing Engineering

    2005-02-02

    Experiments were carried out in a downer reactor integrated in a circulating fluidized bed combustor to examine the performance of the coal topping process. The effects of reaction temperature and coal particle size on the product distribution and their compositions were determined. The experimental results show that an increase in temperature will increase the yields of gas and liquid product, and the liquid yield decreases with the increase in coal particle size. The experiments exhibit an optimal condition for the liquid product. When the pyrolysis temperature is 660{sup o}C and coal particle size is less than 0.2-8 mm, the yield of light tar (hexane-soluble fraction) reaches 7.5 wt % (dry coal basis). The light tar is composed of acid groups (57.1 wt %), crude gasoline (aliphatics) (12.9 wt %), aromatics (21.4 wt %), and polar and basic groups (8.6 wt %). The experiments indicate that the coal topping process is a promising technology for partially converting coal into liquid fuels and fine chemicals.

  14. Design study of toroidal magnets for tokamak experimental power reactors

    International Nuclear Information System (INIS)

    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

  15. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    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 H2O. 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

  16. General Atomic Company fusion experimental power reactor conceptual design

    International Nuclear Information System (INIS)

    The results of a two-year, conceptual design study of a fusion experimental power reactor (EPR) are presented. For this study, the primary objectives of the EPR are to obtain plasma ignition conditions and produce net electrical power. The design features a Doublet plasma configuration with a major radius of 4.5 meters. The average plasma beta is 10 percent which yields a thermonuclear power level of 410 MW during a 105 second burn period. With a duty factor of 0.84, the gross electrical output is 124 MW(e) while the net output is 37 MW(e). The design features a 25 cm thick, helium cooled, modular, stainless steel blanket with a 1 cm thick, thermal radiation-cooled silicon carbide first wall. Sufficient shielding is provided to permit contact maintenance outside the shield envelop within 24 hours after shutdown. An overall facility concept was developed, including a superheated steam cycle power conversion system. Preliminary cost estimates and construction schedules were also developed

  17. Nuclear analysis of a tokamak experimental power reactor conceptual design

    International Nuclear Information System (INIS)

    Detailed nuclear analysis of a reference conceptual design for a tokamak experimental power reactor (EPR) is presented. The reference EPR has a 6.25-m major radius and a 2.1-m minor radius circular plasma with a nominal neutron wall loading of 0.5 MW/m2. A 0.28-m-thick blanket of stainless steel surrounds a stainless-steel vacuum vessel. The inner shield consists of stainless steel and B4C and is 0.58 m thick. The 0.97-m-thick outer shield employs lead mortar, stainless steel, and graphite. The neutronics results in the first wall and blanket vary significantly in the poloidal direction due to an outward shift in the deuterium-tritium neutron source distribution and the toroidal curvature. The infinite cylinder approximation overestimates response rates in the first wall compared with toroidal geometry calculations. Neutral beam lines, vacuum ducts, and other penetrations of the blanket and bulk shield represent large (approximately 0.6- to 1.0-m2 cross section) streaming paths for neutrons and require special shielding. A special 0.75-m-thick annular shield surrounds the neutral beam duct after it exists from the bulk shield and extends beyond the toroidal field coil and out to the beam injectors. A pneumatically operated movable shield plug, opening during the pumpdown phase and closing during the plasma burn, is selected as the principal design option for shielding the evacuation ducts

  18. Conceptual design study of fusion experimental reactor (FY86FER)

    International Nuclear Information System (INIS)

    This report describes the results of the capacity estimation for the electrical power system on the typical two candidates for the FER (Fusion Experimental Reactor) which were picked out through the process of '86 FER scoping studies. Main concern in the electrical systems is coil power supplies which have a capacity of about 1 GW, and this is dominated by poloidal coil power supplies. Then, studies to reduce the converter capacity are concentrated on the poloidal coil power system in relation to the sypplying poloidal flux at the initial phase of plasma ramp-up. A quench protection circuit was proposed on the toroidal coil power supply. On the position control power supply, a circuit with reasonable functions was proposed. Under these system studies, general specifications were determined and the capacity of each power supply unit was estimated. On the poloidal coil power supply system, the accumulated capacity of converters amounted to 885 MW for the one candidate and 782 MW for another. (author)

  19. Poloidal flux linkage requirements for the International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jardin, S.C.; Kessel, C.; Pomphrey, N.

    1994-01-01

    We have applied two computational models to calculate the poloidal flux linkage requirements for the current ramp-up and for the flattop phase of the proposed International Thermonuclear Experimental Reactor (ITER). For the current ramp-up phase, we have used the TSC code to simulate the entire current ramp-up period as described in the TAC-3 Physics Report. We have extended the time of the simulation to cover the full current penetration time, that is, until the loop voltage is a constant throughout the plasma. Sensitivity studies have been performed with respect to current ramp-up time, impurity concentration, and to the time of onset of auxiliary heating. We have also used a steady state plasma equilibrium code that has the constant loop voltage constraint built in to survey the dependence of the steady state loop-voltage on the density and temperature profiles. This calculation takes into account the plasma bootstrap current contribution, including non-circular and collisional corrections. The results can be displayed as contours of the loop-voltage on a POPCON like diagram.

  20. Experimental Study on Granular Sludge Cultivation Based on IC Reactor

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    [Objective]We aimed to discuss the optimal conditions of sludge granulation on the basis of IC reactor.[Method]By using the formulated glucose wastewater,we studied the rapid cultivation of granular sludge as well as its influencing factors,so as to discuss the optimal conditions of sludge granulation.[Result]Through the static culture outside of IC reactor and dynamic culture in IC reactor,granular sludge appeared within a training cycle,with particle size of 1.0-1.5 mm.In IC reactor,when COD concentration...

  1. An option for the Brazilian nuclear project: necessity of fast breeder reactors and core design for an experimental fast reactor

    International Nuclear Information System (INIS)

    Aiming to assure the continued utilization of fission energy, the development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed, as the best type for the Brazilian nuclear system in the future. The inherent safety characteristics are superior to current fast breeder reactors and an efficient utilization of thorium can be realized. The construction and operation of an experimental fast reactor is the first step and a basic tool for the development of FBRs technologies. A serie of core design for an 90 MW FBR is studied and the possible options and sizes of the main parameters are identified. (E.G.)

  2. 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

  3. Magnetic divertors for experimental Tokamaks and fusion reactors

    International Nuclear Information System (INIS)

    Brief reports of working group discussions. These covered the requirements for a divertor in a fusion reactor including reducing impurities, exhausting the plasma and controlling the plasma-wall interactions. Divertor configurations were also reviewed and their merits and disadvantages compared. Existing divertor experiments were summarised and recommendations for further work made. Then the problems anticipated in designing a divertor for a conceptual reactor were considered. The physics of divertors and the scrape-off layer was discussed with reference to present models of plasma in divertors. Finally, experiments needed to demonstrate the feasibility of divertors for reactors and the development of specialised diagnostics for such experiments were considered. (U.K.)

  4. Experimental loop in the Nuclear Training Reactor Budapest

    Energy Technology Data Exchange (ETDEWEB)

    Csom, Gy.; Kocsis, E.; Zsolnay, E.M.; Szondi, E.J.; Szuecs, I. (Budapesti Mueszaki Egyetem (Hungary). Egyetemi Reaktor)

    1982-01-01

    The in-pile loop built into the Nuclear Training Reactor of the Technical University Budapest constructed jointly with the specialists of the Moscow Energetic Institute is used for thermoradiolitic investigations of irradiated solutions under conditions of 20-300 deg C temperature and max. 150 bar pressure. Therefore, the results of such experiments can provide valuable information on the kinetics and mechanisms of chemical processes occurring in the primary and secondary circuits of WWER-type power reactors. In order to obtain results applicable also to power plant conditions, the dose rate had to be increased. Therefore, some modifications of the reactor power were necessary. Preliminary test results are summarized.

  5. 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

  6. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  7. 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.

  8. Teaching Laminar-flow reactors: From experimentation to CFD simulation

    OpenAIRE

    Madeira, LM; Mendes, A.; Magalhaes, FD

    2006-01-01

    An integrated chemical engineering lab experiment is described in this paper. It makes use of a laminar-flow tubular reactor (LFTR) through consecutive lab sessions. In a first session (not described here), the pseudo first-order kinetic constant for the reaction between crystal violet and sodium hydroxide is determined at different temperatures in a batch reactor. Then a tracer experiment is used to characterize the flow, pattern in the LFTR, and finally the steady-state conversion of crysta...

  9. Safety review and assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    More operational events were occurred at various research reactors in 1995. The NNSA and its regional offices conducted careful investigation and strict regulation. In order to analyze comprehensively the safety situation of inservice research reactors and find same countermeasures the NNSA convened a meeting of the safety regulation on research reactors and a meeting for change experience of the safety regulation on research reactors that were participated in by the operating organizations in 1995. A lot of work has been done in the respects of propagation of regulations on nuclear safety, education of nuclear safety culture, the investigation and treatment of operational events, the reexamine of operation documents, the implementation of rectifying items on nuclear safety, the daily inspection and routine inspection on nuclear safety and the studying on the extending service life of research reactors etc

  10. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    International Nuclear Information System (INIS)

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam

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

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; Sint Annaland, van 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 a

  12. Selective catalytic reduction of NO in a reverse-flow reactor: Modelling and experimental validation

    International Nuclear Information System (INIS)

    Highlights: • Reverse-flow reactors easily overcome feed concentration disturbances. • Central feeding improves ammonia adsorption in reverse-flow reactors. • Dynamic heterogeneous model validated with bench-scale experiments. • Optimum reverse-flow reactor design improves efficiency and reduces reactor size. - Abstract: The abatement of nitrogen oxides produced in combustion processes and in the chemical industry requires efficient and reliable technologies capable of fulfilling strict environmental regulations. Selective catalytic reduction (SCR) with ammonia in fixed-bed (monolithic) reactors has stood out among other techniques in the last decades. In this work, the use of reverse-flow reactors, operated under the forced un-steady state generated by the periodic reversal of the flow direction, is studied for improving the SCR performance. This reactor can take advantage of ammonia adsorption in the catalyst to enhance concentration profiles in the reactor, increasing reaction rate, efficiency and reducing the emission of un-reacted ammonia. The process has been studied experimentally in a bench-scale device using a commercial monolithic catalyst. The optimum operating conditions, best ammonia feed configuration (side or central) and capacity of the reactor to deal with feed concentration disturbances is analysed. The experiments have also been used for validating a mathematical model of the reactor based on mass conservation equations, and the model has been used to design a full-size reverse-flow reactor able of operating at industrial conditions

  13. 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

  14. An experimental study of a VVER reactor's steam generator model operating in the condensing mode

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from an experimental study of a VVER reactor's steam generator model operating in the condensing mode are presented. The obtained empirical dependence for calculating the power of heat exchangers operating in the steam condensation mode is presented.

  15. Conceptual design study of fusion experimental reactor (FY 86 FER)

    International Nuclear Information System (INIS)

    This report describes the results of the investigation on critical issues of FY 86 FER reactor configuration/structure design. Accuracy evaluation of shielding calculation and crack growth prediction of first wall and divertor based on the elastic-plastic fracture mechanics were performed. Further, optimization of shield configuration, graphite first wall armor and flexifility of reactor were investigated to support future design work. Feasibilities of innovative ideas were also examined, such as the ripple insert effect and the application of shape memory alloys. (author)

  16. Restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The experimental fast reactor Joyo is the first sodium cooled fast reactor in Japan. Joyo attained initial criticality as a breeder core in April 1977 and has operated as a high performance irradiation test bed since 2003. The 15th periodic inspection of Joyo commenced in May 2007 with the Fuel Handling Machine (FHM) being set up on the Rotating Plug (R/P) for refueling in June. When the R/P was taken down, measuring the load of the Hold-Down Shaft (HDS) revealed an abnormal decrease above the in-vessel storage rack (IVS). The HDS is a cylindrical FMH device that holds down the 6 surrounding subassemblies (S/As) which are adjacent to a withdrawn S/A. In order to investigate the cause of this, an in-vessel observation was conducted using a radiation-resistant fiber scope (RRF). As a result of the observations, it was discovered that the top of the irradiation test S/A 'MARICO-2' (the material testing rig with temperature control) had bent onto the IVS as an obstacle, and had damaged the Upper Core Structure (UCS). During the investigation of this incident, the in-vessel observations using RRF etc. took place at (1) the top of the S/As and the IVS for foreign material, (2) the bottom face of the UCS for damage under the condition with the level of sodium at -50 mm below the top of the S/As. In-vessel observation techniques for a Sodium cooled Fast Reactor (SFR) are important in confirming its safety and integrity. Since an in-vessel observation for an SFR has to be conducted under severe conditions that include high temperatures (∼ 200 deg-C) and high radiation doses (∼ 400 Gy/h), and the primary sodium coolant has to be retained in the Reactor Vessel (R/V) to remove the decay heat, an in-vessel observation equipment has to be designed to not only tolerate the severe conditions but also be capable of being inserted into the sealed R/V through the fixed holes built in to the R/P and gain access to the observation areas. The in-vessel observations were successfully

  17. 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.

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

    International Nuclear Information System (INIS)

    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

  19. European community light water reactor safety research projects. Experimental issue

    International Nuclear Information System (INIS)

    Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

  20. Experimental studies on heat transfer in external cooling of the reactor pressure vessel

    International Nuclear Information System (INIS)

    The filling of the reactor cavity by accidental initiation of the containment spray system, while reactor is in full power, could have severe consequences. If the relatively cold water suddenly cools down the wall of fully pressurized reactor pressure vessel and a crack is assumed to be located on the outer surface of the vessel, the induced thermal stresses might damage the pressure vessel wall. The effects of the inadvertent cooling and pressurized thermal shock (PTS) were studied experimentally at Lappeenranta University of Technology and the heat transfer coefficients gained from the experimental results were compared with calculations. (author)

  1. Stable hydrogen production by methane steam reforming in a two zone fluidized bed reactor: Experimental assessment

    Science.gov (United States)

    Pérez-Moreno, L.; Soler, J.; Herguido, J.; Menéndez, M.

    2013-12-01

    The Two Zone Fluidized Bed Reactor concept is proposed for hydrogen production via the steam reforming of methane (SRM) including integrated catalyst regeneration. In order to study the effect of the contact mode, the oxidative SRM has been carried out over a Ni/Al2O3 catalyst using a fixed bed reactor (fBR), a conventional fluidized-bed reactor (FBR) and the proposed two-zone fluidized bed reactor (TZFBR). The technical feasibility of these reactors has been studied experimentally, investigating their performance (CH4 conversion, CO and H2 selectivity, and H2 global yield) and stability under different operating conditions. Coke generation in the process has been verified by several techniques. A stable performance was obtained in the TZFBR, where coke formation was counteracted with continuous catalyst regeneration. The viability of the TZFBR for carrying out this process with a valuable global yield to hydrogen is demonstrated.

  2. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    International Nuclear Information System (INIS)

    The Oak Ridge National Laboratory has three multipurpose research reactors which accommodate testing loops, target irradiations, and beam-type experiments. Since the experiments must share common or similar facilities and utilities, be designed and fabricated by the same groups, and meet the same safety criteria, certain standards for these have been developed. These standards deal only with those properties from which safety and economy of time and money can be maximized and do not relate to the intent of the experiment or quality of the data obtained. The necessity for, and the limitations of, the standards are discussed; and a compilation of general standards is included

  3. Experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core

    International Nuclear Information System (INIS)

    The experimental study of prompt neutron decay constant α for 300# pool reactor under mixed core was carried out through a suit of reactor power spectral density measurement system. The two channel continuous current signals of neutron in the reactor were acquired by ionization chamber DL129 which was symmetrically putted in reactor core. The power spectral density, for two channel signals, was computed using the application program of data acquirement and data process analysis. Finally, by using the non-linear least squares method, the prompt neutron decay constant α was fitted. By comparison, the experimental results well accord to the theory calculation within the error range. The deviation can meet the actual need of project. (authors)

  4. 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.

  5. 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)

  6. Reflooding of a severely damaged reactor core. Experimental analysis and modelling

    International Nuclear Information System (INIS)

    The understanding of the reflood process of a severely damaged reactor core represents a challenge in the prediction of safety margin of existing and future pressurized water reactors. After the TMI-2 accident, the understanding of coolability of severely damaged reactor core became an objective of many theoretical and experimental studies. Currently, the French Institute of Radioprotection and Nuclear Safety (IRSN) has started two experimental programs, PRELUDE and PEARL, to investigate the physical phenomena during a reflood process at high temperature and to provide relevant data in order to improve predictive models. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core. The presented model is based on the theory of heat transfer and two-phase flow in porous media and in small hydraulic diameter channels. The proposed model is implemented into the European computer code for severe accident analysis ICARE-CATHARE. The comparison of the calculations with PRELUDE experimental results is presented. Finally, the issue of transposition to the reactor scale is discussed and some answers are proposed using calculation results for a debris bed in a configuration similar to what could be expected in a severely damaged reactor core. (author)

  7. 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.

  8. 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...... the experimental uncertainty. The residual limestone level in the gypsum formed increased with increasing values of reactor slurry pH, inlet flue gas SO2 concentration, and slurry concentration of Cl-. (C) 2003 Elsevier Ltd. All rights reserved....

  9. 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

  10. Studies on plasma shutdown of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Shutdown of the plasma with a time-dependent one-point model is described. The pseudoclassical scaling law plays a role in the plasma diffusion in the low energy region below several keV and the trapped ion scaling law in the higher energy region. In this shutdown model, only deuterium is inserted during 20-second shutdown process. In the first 10 sec, while the plasma temperature, electron density and plasma current decrease from 7 keV to 1 keV, 1.1 x 1020m-3 to 1019m-3 and 4 MA respectively the fusion power falls down with gradual decrease of heating power. During the second 10 sec, while the plasma temperature, electron density and plasma current decrease from 1 keV to 100 keV, 1019m-3 to 1018m-3 and 1 MA to 100 kA respectively, the plasma thermal energy is removed. Plasma one-turn voltages are -4.0 volt and -0.5 -- -1.0 volt which fall the plasma current down to 1 MA and 100 kA during the first 10 sec and the second 10 sec, respectively. Decrease of plasma current largely lowers plasma density and energy since particle and energy confinement times decrease as plasma current decreases. Deuterium insertion rate below that in the equilibrium operation little lowers plasma density and energy. This plasma shutdown scheme is effective in driven-type reactors. (auth.)

  11. Conceptual design study of fusion experimental reactor (FY86 FER)

    International Nuclear Information System (INIS)

    This report summarizes the FER magnet design which was conducted last year (1986). Main objective of the new FER design is to have better cost performance of the machine. The physics assumptions are reviewed to reduce risks. Optimization of the physics design and improvements of the engineering design have been done without changing missions of the device. After a preliminary investigation for the optimization and improvements, six FER concepts have been developed to establish the improved design point, and have been studied in more detail. In the magnet design, the improvements of superconducting magnet design were mainly investigated to reduce the reactor size. A normal conductor was studied as an alternative option for appling to the special poloidal field coils that were located on the interior to the toroidal field coils. Some improvements were made on the superconducting magnet design. Based on the preliminary investigation, the magnet design specifications have been modified somewhat. The conceptual design of the magnet system components have been done for the candidate FER concepts. (author)

  12. CO2 Absorption in a Lab-Scale Fixed Solid Bed Reactor: Modelling and Experimental Tests

    Directory of Open Access Journals (Sweden)

    Roberto Gabbrielli

    2004-09-01

    Full Text Available The CO2 absorption in a lab-scale fixed solid bed reactor filled with different solid sorbents has been studied under different operative conditions regarding temperature (20-200°C and input gas composition (N2, O2, CO2, H2O at 1bar pressure. The gas leaving the reactor has been analysed to measure the CO2 and O2 concentrations and, consequently, to evaluate the overall CO2 removal efficiency. In order to study the influence of solid sorbent type (i.e. CaO, coal bottom ash, limestone and blast furnace slag and of mass and heat transfer processes on CO2 removal efficiency, a one-dimensional time dependent mathematical model of the reactor, which may be considered a Plug Flow Reactor, has been developed. The quality of the model has been confirmed using the experimental results.

  13. Radiological characteristics of light-water reactor spent fuel: A literature survey of experimental data

    International Nuclear Information System (INIS)

    This survey brings together the experimentally determined light-water reactor spent fuel data comprising radionuclide composition, decay heat, and photon and neutron generation rates as identified in a literature survey. Many citations compare these data with values calculated using a radionuclide generation and depletion computer code, ORIGEN, and these comparisons have been included. ORIGEN is a widely recognized method for estimating the actinide, fission product, and activation product contents of irradiated reactor fuel, as well as the resulting heat generation and radiation levels. These estimates are used as source terms in safety evaluations of operating reactors, for evaluation of fuel behavior and regulation of the at-reactor storage, for transportation studies, and for evaluation of the ultimate geologic storage of spent fuel. 82 refs., 4 figs., 17 tabs

  14. 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.

  15. Experimental determination of neutron capture cross sections of fast reactor structure materials integrated in intermediate energy spectra. Vol. 2: description of experimental structure

    International Nuclear Information System (INIS)

    A selection of technical documents is given concerning the experimental determination of the neutron capture cross-sections of fast reactor structural materials (Fe, Cr, Ni...) integrated over the intermediate energy spectra. The experimental structure project and modifications of the reactor RB2 for this experiment, together with criticality and safety calculations, are presented

  16. Experimental study on thermal stratification in a reactor hot plenum of a Japanese demonstration LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Koga, Tomonari [Central Research Inst. of Electric Power Industry, Abiko, Chiba (Japan). Abiko Research Lab.; Yamamoto, K.; Takakuwa, M.; Kajiwara, H.; Watanabe, O.; Akamatsu, K.

    1997-12-31

    Thermal stratification which occurs in a reactor hot plenum after reactor trip has been regarded as one of the most serious phenomena in the thermal-hydraulics of LMFBR. Using a 1/8th scale water model, an experimental study has been conducted to estimate the thermal stratification for a Japanese demonstration LMFBR (DFBR). In the present study, reactor trip was simulated by changing the core outlet temperature with maintaining a constant flow rate. Temperature distribution was measured during the transient and detailed phenomena have been acquired in the study. A severe density interface on structural integrity occurs in a hot plenum under the thermal stratification. Experimental results for temperature gradient and rising speed of the density interface were estimated based on a similarity rule so that an actual condition in the DFBR could be fully discerned. (author)

  17. Further optimization studies of experimental dynamic responses measured on the HTGC Dragon reactor

    International Nuclear Information System (INIS)

    This report considers some measurements made of the dynamics of the HTGC Dragon reactor and the optimization of a mathematical model which represents the reactor, by altering the parameters until a least squares fit between the experimental responses and the mathematical model is obtained. The experimental information was processed in various ways. The experimental response to an impulse, step or periodic sine wave change in reactivity was processed as an impulse, step or periodic sine wave response respectively and compared with a similar response from the model. In other studies the result of a binary cross correlation experiment (effectively an impulse response input) was processed as a frequency response and this experimental frequency response was compared with the frequency response from the mathematical model. It was possible therefore to compare the optimum values of parameters, obtained for different forms of perturbing signal and for different methods of processing and to relate the optima obtained to the problem of parameter estimation. (author)

  18. 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...

  19. Jules Horowitz reactor. France development of an experimental loop integrating an optimized irradiation process

    International Nuclear Information System (INIS)

    Experimental reactors in the world enable researchers to meet the needs of industry and institutions not only by providing support to the existing nuclear infrastructure (Gen.2) but also by preparing the future generation (Gen.3, Gen.4) or even by responding to other needs as well (supports with fusion, medical applications). It is for this specific purpose that the Jules Horowitz Polyvalent Irradiation Reactor is now being built at the CEA Cadarache Research Center (located in the south of France). This Material Testing Reactor (MTR type) is designed to irradiate materials or fuel samples for various experimental tests. The reactor will also produce Mo99 radioelements that will supply 25% to 50% of current European needs. The goal of this paper is to describe a fuel irradiation loop, now under study, that will be designed to carry out power ramps tests to provide technical support to the Generation 2 and 3 nuclear power plants. In order to increase its irradiation capacity (2 to 3 per cycle), this loop takes into account the requirements that will lead to the optimization of all experimental processes in the facility (such as non-destructive examinations before and after the test, specific loading tools). All these considerations are being taken into account in order to offer to the customer's complete and optimized conditions in terms of experimental irradiations processes. (author)

  20. Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

    2013-05-06

    The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

  1. Membrane assisted fluidized bed reactor: experimental demonstration for partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, Salim Abdul Rashid Khan

    2004-01-01

    In this thesis the reactor concept has been developed on the basis of an experimental study on the effect of fluidization conditions on the membrane permeation rate in a MAFBR, the extent of gas back mixing and the tube-to-bed heat transfer rates in the presence of membrane bundles with and without

  2. Students' assessment of interactive distance experimentation in nuclear reactor physics laboratory education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-10-01

    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 reactors limit their accessibility to few educational programmes around the world. The concept of the Internet Reactor Laboratory (IRL) was introduced earlier as a new approach that utilises distance education in nuclear reactor physics laboratory education. This paper presents an initial assessment of the implementation of the IRL between the PULSTAR research reactor at North Carolina State University in the USA and the Department of Nuclear Engineering at Jordan University of Science and Technology (JUST) in Jordan. The IRL was implemented in teaching the Nuclear Reactor laboratory course for two semesters. Feedback from surveyed students verifies that the outcomes attained from using IRL in experimentation are comparable to that attainable from other on-campus laboratories performed by the students.

  3. 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.

  4. 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)

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

    International Nuclear Information System (INIS)

    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)

  6. 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.

  7. Operational experience and upgrading program of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Twenty years of successful operations at the experimental fast reactor JOYO provide a wealth of experience covering core management, chemical analysis of sodium and cover gas for impurity control, natural convection tests, upgrade of fuel failure detection system, corrosion product measurement, development of operation and maintenance support system, and replacement of major components in the cooling systems. Some of the data obtained is stored in a database to preserve the related knowledge. This experience and accumulated data will be useful for the design of future fast reactors. (author)

  8. Development of a remote handling system for replacement of armor tiles in the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    The armor tiles of the Fusion Experimental Reactor (FER) planned by JAERI are categorized as scheduled maintenance components, since they are damaged by severe heat and particle loads from the plasma during operation. A remote handling system is thus required to replace a large number of tiles rapidly in the highly activated reactor. However, the simple teaching-playback method cannot be adapted to this system because of deflection of the tiles caused by thermal deformation and so on. We have developed a control system using visual feedback control to adapt to this deflection and an end-effector for a single arm. We confirm their performance in tests. (orig.)

  9. Two dimensional neutron transport calculation system for plate-reactors: experimental design and qualification with SILOE

    International Nuclear Information System (INIS)

    The main objective of this work is to create a neutronic calculations system for the SILOE-SILOETTE reactors, adaptable to other types of plate reactors. The author presents the methodology and the development of the APOLLO 1D (99 gr.) calculations for the creation of cross sections libraries. After a recall of the Discrete Ordinate Method (DOT), the method accuracy is studied in order to optimize the spatial discretization of the calculations; calculations of DOT 3.5 and of SILOETTE core are conducted and their convergence and costs are examined. DOT calculations of SILOETTE and experimental tests results are then compared

  10. An Experimental Study of Natural Convection in The Hottest Channel of TRIGA 2000 k W Reactor

    International Nuclear Information System (INIS)

    With the increase of radioisotope demand, in 1995, National Nuclear Energy Agency of Indonesia made a decision to upgrade the power of the TRIGA Mark II reactor from 1 MW to 2 MW maximum power. The reactor reached its first criticality on May 13, 2000. To accomplish the safety evaluation of the reactor, a thermal hydraulic analysis was carried out by using thermal hydraulic computer code. This code calculates the natural convection flow through water coolant bounded by vertical cylindrical heat sources. In this paper, it will be reported the experimental study of natural convection in the hottest channel of TRIGA 2000 k W reactor. The purpose of the experimental study is to verify the theoretical analysis, especially the temperature distribution in the hottest coolant channel. In this experiment, a special probe for temperature detection has been designed and inserted to central thimble (CT). In the experiment, eight thermocouples were used to measure the bulk temperature of the water at different position in the cooling channel and simultaneous quantitative measurement of the temperature distribution were done by using a data acquisition cards system. The result obtained theoretically using the STAT code has been verified by this experimental study. (author)

  11. Implementation of multivariable control techniques with application to Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    After several successful applications to aerospace industry, the modern control theory methods have recently attracted many control engineers from other engineering disciplines. For advanced nuclear reactors, the modern control theory may provide major advantages in safety, availability, and economic aspects. This report is intended to illustrate the feasibility of applying the linear quadratic Gaussian (LQG) compensator in nuclear reactor applications. The LQG design is compared with the existing classical control schemes. Both approaches are tested using the Experimental Breeder Reactor 2 (EBR-2) as the system. The experiments are performed using a mathematical model of the EBR-2 plant. Despite the fact that the controller and plant models do not include all known physical constraints, the results are encouraging. This preliminary study provides an informative, introductory picture for future considerations of using modern control theory methods in nuclear industry. 10 refs., 25 figs

  12. Perturbation method for experimental determination of neutron spatial distribution in the reactor cell

    International Nuclear Information System (INIS)

    The method is based on perturbation of the reactor cell from a few up to few tens of percent. Measurements were performed for square lattice calls of zero power reactors Anna, NORA and RB, with metal uranium and uranium oxide fuel elements, water, heavy water and graphite moderators. Character and functional dependence of perturbations were obtained from the experimental results. Zero perturbation was determined by extrapolation thus obtaining the real physical neutron flux distribution in the reactor cell. Simple diffusion theory for partial plate cell perturbation was developed for verification of the perturbation method. The results of these calculation proved that introducing the perturbation sample in the fuel results in flattening the thermal neutron density dependent on the amplitude of the applied perturbation. Extrapolation applied for perturbed distributions was found to be justified

  13. EXPERIMENTAL STUDY OF THYRISTOR CONTROLLED REACTOR (TCR AND GTO CONTROLLED SERIES CAPACITOR (GCSC

    Directory of Open Access Journals (Sweden)

    JYOTI AGRAWAL,

    2011-06-01

    Full Text Available This paper deals with the simulation of Thyristor controlled reactor (TCR and GTO Controlled Series Capacitor (GCSC, equipment for controlled series compensation of transmission systems. The paper alsopresents experimental results of a TCR and GCSC connected to a single-phase system. The experiments are carried out in the FACTS lab of electrical engineering department. The TCR system is simulated using MATLAB and the simulation results are presented. The power and control circuits are simulated. The current drawn by the TCR varies with the variation in the firing angle. Stepped variation of current can be obtained using thyristor switched reactor. The simulation results are compared with the theoretical and practical results.Harmonics and its impact on the system are presented. This paper also presents the GCSC, its main components, principal of operation, typical waveforms and main applications. Duality of the GCSC with the well known thyristor controlled reactor is also discussed in this paper.

  14. Development, Implementation and Experimental Validations of Activation Products Models for Water Pool Reactors

    International Nuclear Information System (INIS)

    Some parameters were obtained both calculations and experiments in order to determined the source of the meaning activation products in water pool reactors. In this case, the study was done in RA-6 reactor (Centro Atomico Bariloche - Argentina).In normal operation, neutron flux on core activates aluminium plates.The activity on coolant water came from its impurities activation and meanly from some quantity of aluminium that, once activated, leave the cladding and is transported by water cooling system.This quantity depends of the 'recoil range' of each activation reaction.The 'staying time' on pool (the time that nuclides are circulating on the reactor pool) is another characteristic parameter of the system.Stationary state activity of some nuclides depends of this time.Also, several theoretical models of activation on coolant water system are showed, and their experimental validations

  15. Experimental methods of investigation of kinetics and dynamics of nuclear reactors

    International Nuclear Information System (INIS)

    The author presents experimental methods used to study kinetic and dynamic properties of nuclear reactors. Kinetic methods aim at determining characteristic parameters of the behaviour in time of neutrons. Dynamic methods aim at establishing the relationships between the reactor behaviour and its internal and external causes (notably the measurement of transfer functions). The author proposes a classification with respect to the excitation type: periodic excitation (reactivity sinusoidal modulation, source sinusoidal modulation, periodic pulse excitation), non periodic excitation (reactivity monitoring, reactivity linear variation, reactivity variation according to any given law, removal of starting source), random excitation (random reactivity or source excitation), natural fluctuations (alpha-Rossi method, methods of reduced variance, probabilistic methods, correlation methods, spectral analysis method). He also addresses space and energy effects. Applications are reported for low power and power reactors

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

    International Nuclear Information System (INIS)

    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

  17. Neutronic and thermal-hydraulic experimental program in the IPR-R1 TRIGA reactor at CDTN

    International Nuclear Information System (INIS)

    The IPR-R1 TRIGA reactor, located at CDTN (Belo Horizonte/Brazil), is a typical 100 kW Mark I light-water reactor cooled by assisted natural convection with an annular graphite reflector. In order to study the safety aspects connected with the increase of the maximum steady state power of the IPR-R1 TRIGA reactor, experimental measures were taken. This paper summarizes the experimental program and some recent results and procedures of the neutronic and thermalhydraulic experiments carried out in the IPR-R1 TRIGA reactor. (authors)

  18. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  19. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  20. Experimental and modeling study of sulfur dioxide oxidation in packed-bed tubular reactor

    Directory of Open Access Journals (Sweden)

    Hanen NOURI

    2013-08-01

    Full Text Available The conversion of sulfur dioxide into sulfur trioxide is a reaction which interests not only the industry of sulfuric acid production but also the processes of pollution control of certain gas effluents containing SO2. This exothermic reaction needs a very good control of temperature, that's why it is led in the industry in a multistage converter with intermediate heat exchangers. Microreactors represent a good alternative for such reaction due to their intensification of mass and heat transfer and enhancement of temperature control. In this study, this reaction was conducted in a stainless steel tubular (4mm ID packed bed reactor using particles of vanadium pentoxide as catalyst at atmospheric pressure. Experiments were performed with different inlet SO2 concentration in 3-9% range and reaction temperature between 685-833K. We noticed that the conversion decreases with the amount of SO2 and increases with the temperature until an optimum, above this value the conversion drop according to the shape of the equilibrium curve. Controlling rate mechanism is studied by varying temperature. Pseudohomogeneous perfect plug flow is used to describe this small tubular reactor. Numerical simulations with MATLAB were performed to validate the experimental results. Good agreement between the model predictions and the experimental results is achieved. Fluid flow description inside the packed bed reactor was performed by using the free fluid and porous media flow model. This model was solved by the commercial software COMSOL Multiphysics. Velocity profile inside the reactor is theoretically obtained.

  1. Experimental studies of the oscillatory combustion of hydrogen in a stirred flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baulch, D.L.; Griffiths, J.F. (Leeds Univ. (United Kingdom). School of Chemistry); Kordylewski, W. (Technical Univ. of Wroclaw (PL). Dept. of Heat Engineering and Fluid Mechanics); Richter, R. (Southampton Univ. (United Kingdom). Dept. of Chemistry)

    1991-11-15

    Experimental studies of the phase relations between H atoms. OH radicals and reactant temperature during the gas-phase, oscillatory combustion of hydrogen in a well-stirred flow reactor are reported. Absolute concentrations of the OH radical and the reactant temperature were measured in absorption from the vibrational rotation structure of the laser-induced, electronically excited, OH spectrum. Relative concentrations of H atoms were obtained by multiphoton ionization, also induced by a laser. The hydrogen atoms reached their maximum concentration first during the oscillatory combustion rising to a sharp peak followed by a rapid decay within several milliseconds. The OH radicals reached their maximum concentration about 1 ms after the H atoms. The maximum of the reactant temperature was in phase with the hydroxyl radicals. Experimental and numerical studies of the interaction that occurs between oscillations in a pair of coupled reactors are also presented. (author).

  2. 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 safe......, (b) process monitoring and control and (c) man-machine communications. The experimental system configuration is described and relevant hardware and software matters discussed in some detail...

  3. 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)

  4. BOR-60 reactor as an instrument for experimental substantiation of fuel rods for advanced NPPs

    International Nuclear Information System (INIS)

    Full text: The BOR-60 fast test reactor is actually the only facility of this type in the world that has been in reliable and continuous operation for about 35 years. One of the principle reactor tasks is irradiation of advanced fuel and structural materials in different conditions. Inside the reactor the materials can be irradiated in any core and reflector cell except seven cells used for control rods. The number of fuel assemblies loaded into the reactor can vary from 85 to 124 depending on the burnup, core configuration and fuel properties. Due to the reactor design, the core dimensions can be widely changed allowing accommodation of no less than 20 experimental assemblies in different reactor cells. The neutron flux value in individual cells can vary more than 3 times at the maximum value of 3.7·1015 n/cm2s. Thus various fuel compositions can be loaded into the reactor and practically any burnups can achieve. Based on the long-term investigation of the reactor characteristics, we studied the reactor behavior in different conditions, developed a set of the verified codes and different procedures for the on-line reactor maintenance and performance of the wide scope of experiments. A set of specialized testing facilities consisting of capsule units and dismountable assemblies are used for irradiation of the wide range of materials and items at different conditions. The advantages of these facilities are their simplicity and possibility of installation in any core and reflector cell. In addition to the precision calculations of the irradiation conditions there is also a possibility for monitoring the neutron flux and temperature. A special thermometric channel available in the core allows accommodation of the experimental facilities and output of information of the irradiation conditions by 30-50 communication lines. It was required to develop a series of independent instrumented capsule-loops, special instrumented fuel assemblies etc. to be used in the channel

  5. Experimental techniques to determine salt formation and deposition in supercritical water oxidation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, J.P.C.; LaJeunesse, C.A.; Rice, S.F.

    1994-08-01

    Supercritical Water Oxidation (SCWO) is an emerging technology for destroying aqueous organic waste. Feed material, containing organic waste at concentrations typically less than 10 wt % in water, is pressurized and heated to conditions above water`s critical point where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. An oxidizer, is then added to the feed. Given adequate residence time and reaction temperature, the SCWO process rapidly produces innocuous combustion products. Organic carbon and nitrogen in the feed emerge as CO{sub 2} and N{sub 2}; metals, heteroatoms, and halides appear in the effluent as inorganic salts and acids. The oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms acid anions. In the presence of metal ions, salts are formed and precipitate out of the supercritical fluid. In a tubular configured reactor, these salts agglomerate, adhere to the reactor wall, and eventually interfere by causing a flow restriction in the reactor leading to an increase in pressure. This rapid precipitation is due to an extreme drop in salt solubility that occurs as the feed stream becomes supercritical. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. A phenomenological model is developed in this paper to predict the time that reactor pressure begins to rise as a function of the fluid axial temperature profile and effective solubility curve. The experimental techniques used to generate effective solubility curves for one salt of interest, Na{sub 2}SO{sub 4}, are described, and data is generated for comparison. Good correlation between the model and experiment is shown. An operational technique is also discussed that allows the deposited salt to be redissolved in a single phase and removed from the affected portion of the reactor. This technique is demonstrated experimentally.

  6. Experimental results of angular neutron flux spectra leaking from slabs of fusion reactor candidate materials, (1)

    International Nuclear Information System (INIS)

    This report summarizes experimental data of angular neutron flux spectra measured on the slab assemblies of fusion reactor candidate materials using the neutron time-of-flight (TOF) method. These experiments have been performed for graphite (carbon), beryllium and lithium-oxide. The obtained data are very suitable for the benchmark tests to check the nuclear data and calculational code systems. For use of that purpose, the experimental conditions, definitions of key terms and results obtained are compiled in figures and numerical tables. (author)

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  10. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  11. 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...

  12. Theoretical and experimental studies of fixed-bed coal gasification reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Joseph, B.; Bhattacharya, A.; Salam, L.; Dudukovic, M.P.

    1983-09-01

    A laboratory fixed-bed gasification reactor was designed and built with the objective of collecting operational data for model validation and parameter estimation. The reactor consists of a 4 inch stainless steel tube filled with coal or char. Air and steam is fed at one end of the reactor and the dynamic progress of gasification in the coal or char bed is observed through thermocouples mounted at various radial and axial locations. Product gas compositions are also monitored as a function of time. Results of gasification runs using Wyoming coal are included in this report. In parallel with the experimental study, a two-dimensional model of moving bed gasifiers was developed, coded into a computer program and tested. This model was used to study the laboratory gasifier by setting the coal feed rate equal to zero. The model is based on prior work on steady state and dynamic modeling done at Washington University and published elsewhere in the literature. Comparisons are made between model predictions and experimental results. These are also included in this report. 23 references, 18 figures, 6 tables.

  13. 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.

  14. Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

    International Nuclear Information System (INIS)

    Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li2TiO3, study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC UMP'. Scientific and technical support of these

  15. The Jules Horowitz Reactor : A new high Performances European MTR (Material Testing Reactor) with modern experimental capacities : Toward an International User Facility

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major Research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactor design. It will represent also an important Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which was started in 2007 is on-going. The first operation is planned before the end of this decade.The design of the reactor will provide an essential facility supporting the programs for the nuclear energy for the next 50 years. JHR is designed to provide high neutron flux (enhancing the maximum available today in MTRs), to run highly instrumented experiments to support advanced modelling giving prediction beyond experimental points, and to operate experimental devices giving environment conditions (pressure, temperature, flux, coolant chemistry, ···) relevant for water reactors, for gas cooled thermal or fast reactors, for sodium fast reactors, ···So, the reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation possibilities for future reactors. In parallel to the construction of the reactor, the preparation of an international community around JHR is continuing; this is an important topic as building and gathering a strong international community in support to MTR experiments is a key-issue for the R and D in nuclear energy field. Consequently, CEA is

  16. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Barbot, L.; Domergue, C.; Destouches, C. [CEA DEN, DER, Instrumentation Sensors and Dosimetry laboratory Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  17. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  18. Simulations of alpha particle ripple loss from the International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Redi, M.H.; Budny, R.V.; McCune, D.C.; Miller, C.O.; White, R.B.

    1996-05-01

    Calculations of collisional stochastic ripple loss of alpha particles from the new 20 toroidal field (TF) coil International Thermonuclear Experimental Reactor (ITER) predict small alpha ripple losses, less than 0.4%, close to the loss calculated for the full current operation of the earlier 24 TF coil design. An analytic fit is obtained to the ITER ripple data field demonstrating the nonlinear height dependence of the ripple minimum for D shaped ripple contours. In contrast to alpha loss simulations for the Tokamak Fusion Test Reactor (TFTR), a simple Goldston, White, Boozer stochastic loss criterion ripple loss model is found to require an increased renormalization of the stochastic threshold {delta}{sub s}/{delta}{sub GWB} {ge} 1. Effects of collisions, sawtooth broadening and reversal of the grad B drift direction are included in the particle following simulations.

  19. Simulations of alpha particle ripple loss from the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Calculations of collisional stochastic ripple loss of alpha particles from the new 20 toroidal field (TF) coil International Thermonuclear Experimental Reactor (ITER) predict small alpha ripple losses, less than 0.4%, close to the loss calculated for the full current operation of the earlier 24 TF coil design. An analytic fit is obtained to the ITER ripple data field demonstrating the nonlinear height dependence of the ripple minimum for D shaped ripple contours. In contrast to alpha loss simulations for the Tokamak Fusion Test Reactor (TFTR), a simple Goldston, White, Boozer stochastic loss criterion ripple loss model is found to require an increased renormalization of the stochastic threshold δs/δGWB ≥ 1. Effects of collisions, sawtooth broadening and reversal of the grad B drift direction are included in the particle following simulations

  20. 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

  1. Experimental and analytical studies of high heat flux components for fusion experimental reactor

    International Nuclear Information System (INIS)

    In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 ± 1 MW/m2 was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate has been analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads. (J.P.N.) 62 refs

  2. Three-dimensional simulation and experimental investigation of a novel biomass fast pyrolysis reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H.Y.; Shao, S.S.; Xiao, R.; Pan, Q.W.; Chen, R.; Zhang, J.B. [Southeast Univ., Nanjing (China). School of Energy and Environment

    2013-07-01

    A novel autothermal reactor, named internally interconnected fluidized beds (IIFB), was developed for biomass fast pyrolysis to produce liquid fuels and chemicals. The IIFB reactor includes a pyrolysis bed and a combustion bed to conduct biomass pyrolysis and char burning, respectively. In this study, numerical simulation and experimental studies on volume fraction of particles, solid circulation rate and pressure distribution of the IIFB are reported. The stable flow photographed from the simulations coincides with that in the experiments at the same operating conditions. At the same height, the velocity of gas is twice as larger as the velocity of solid, which is favorable for catalytic reactions. The particles move up unsteadily in the draft tube, and yet they fall down with an almost constant velocity 0.07 m/s in the dipleg. The pressure in the fluidization region is higher than that in the spouted region at H=10mm and it shows an opposite pressure distribution. It is also observed that the experimental value of pressure is in well agreement with that obtained from simulations on the bottom, and yet it shows very different characteristics on the two outlets. Simulation results show that solid circulation rate at different cross-sections converged to 110kg/h which is in well agreement with experimental data of 104.5kg/h.

  3. Influence of operation of national experimental nuclear reactor on the natural environment

    Directory of Open Access Journals (Sweden)

    Agnieszka Kaczmarek-Kacprzak

    2012-09-01

    Full Text Available This paper presents the impact of experimental nuclear reactor operations on the national environment, based on assessment reports of the radiological protection of active nuclear technology sources. Using the analysis of measurements carried out in the last 15 years, the trends are presented in selected elements of the environment on the Świerk Nuclear Centre site and its surroundings. In addition, the impact of research results is presented from the fi fteen year period of environmental analysis on building public confi dence on the eve of the start of construction of the first Polish nuclear power plant.

  4. Processing of uranium oxide powders in a fluidized-bed reactor. I. Experimental

    Science.gov (United States)

    Cho, W. D.; Han, Man-Hee; Bronson, Mark C.; Zundelevich, Yury

    2002-10-01

    The oxidation of UN powders was carried out in a spout-type fluidized-bed reactor in gas mixtures of oxygen and argon, and over the temperature range of 200-500 °C. The rate of the conversion from UN to U 3O 8 powders was measured using gas chromatography and found to be dependent on temperature, partial pressure of oxygen and gas flowrate. The solid reactants and products were analyzed using SEM and XRD. Based on the experimental results, the conversion process was explained by the crackling core model.

  5. An alternative experimental approach for subcritical configurations of the IPEN/MB-01 nuclear reactor

    Science.gov (United States)

    Gonnelli, E.; Lee, S. M.; Pinto, L. N.; Landim, H. R.; Diniz, R.; Jerez, R.; dos Santos, A.

    2015-07-01

    This work presents an alternative approach for the reactivity worth experiments analysis in the IPEN/MB-01 reactor considering highly subcritical arrays. In order to reach the subcritical levels, the removal of a specific number of fuel rods is proposed. Twenty three configurations were carried out for this purpose. The control bank insertion experiment was used only as reference for the fuel rod experiment and, in addition, the control banks were maintained completely withdrawn during all the fuel rods experiment. The theoretical simulation results using the MCNP5 code and the ENDF/B-VII.0 library neutron data are in a very good agreement to experimental results.

  6. Method Experimental Definition of Efficiency Fuelburn in a Gas Reactors with Mini Channels

    Directory of Open Access Journals (Sweden)

    Slesareva Ekaterina

    2016-01-01

    Full Text Available Approbation of a method of experimental definition of efficiency of a fuelburn in minichanneling gas reactors has been carried out. The method of visualisation of a field of temperatures by the temperature transducer made of fine-meshed with low-inertia use for definition of thermal structure of gas streams. A field of temperatures on an outlet the simulator of a gas reactor with minichannels registered on thermogram by infrared imager. Experiments have shown that the thermal imaging method provides the good resolutions on spatial and on time for a field of temperatures both in stationary and in non-stationary regime of gas flow and of heat release. Qualitative and quantitative character a change a temperatures on thermograms correspond to classical physical representations about nature of а gas flow in lengthy channels. The study showed that the method of thermal imaging allows in a regime real time defined a parameters of gas flow on outlet a channels of reactor which are necessary for definition of fuelburn efficiency.

  7. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  8. Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

    International Nuclear Information System (INIS)

    The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

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

    International Nuclear Information System (INIS)

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U235 are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained

  10. 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

  11. Experimental investigation on flow behavior during start-up of a heating reactor

    International Nuclear Information System (INIS)

    An experimental simulation study on the transition from pressurized to boiling operation of a low-temperature, natural circulation nuclear heating reactor (5 MW) developed by INET of Tsinghua University is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and low-steam quality density wave instability on the transition from pressurized to boiling operation is described. The mechanism of flashing instability, which has never been studied well on this field, is especially interpreted. It is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: (1) increasing of initial pressure by means of a noncondensable gas (N2), which is a very effective method to eliminate geyser instability and flashing instability at lower pressure. (2)start-up of the reactor at this pressurized condition with a constant heat flux under the limited value of q = 0.15 MW·m-2, which controls the exit temperature of the heated section below the one of net vapor generation, the low steam quality density wave oscillation can be avoided. (3) transition to a lower pressure, boiling operation. The method of transition with low-heat flux and low-inlet subcooling is proposed: at pressurized operation condition, by reducing the heat flux to its lowest level, releasing the noncondensable gas and increasing the heat flux gradually (dq/dt-2·min-1), during which the low-steam quality density wave oscillation can be prevented from occurring, then the boiling operation condition can be achieved through adjusting the heat flux and inlet subcooling to their designed value. A stable transition from pressurized to boiling operation of the 5 MW reactor is achieved by careful selection of the thermohydraulic parameters. (7 refs., 7 figs., 1 tab.)

  12. 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

  13. The siting of an experimental fusion reactor in Canada: Economic impact and technical feasibility

    International Nuclear Information System (INIS)

    Fusion energy has been heralded as the successor to current nuclear fission technology, and as a long-term environmentally sustainable resource. Since 1988, Canada has been a participant in a project to build the world's first fusion energy engineering test reactor, called ITER (International Thermonuclear Experimental Reactor). The purpose of ITER is to demonstrate the scientific and technical feasibility of magnetic fusion energy. The ITER project represents a first-of-its-kind international collaboration between four major parties, the European Communities, the Russian Federation, Japan, and the United States. At an estimated construction cost of almost $10B (1993 $CDN), it will service as a world focus for an array of leading-edge technologies, with potential applications in almost every major high-tech industry. Canada has participated as a contributor to the European Community effort and currently donates in-kind contributions of advanced technology developed in Canada. The Canadian Fusion Fuels Technology Project (CFFTP) is responsible for providing and coordinating Canadian engineering and technology support for the ITER project, while the Centre canadien de fusion magnetique (CCFM) provides physics support. In March 1993, Wardrop Engineering Inc. was retained by CFFTP to assess the technical feasibility of siting an ITER-tpe experimental fusion reactor at a representative existing nuclear site in Canada, and to conduct a preliminary economic analysis of Canadian economic impacts resulting from a venture of this type. This paper details the results of this work, and addresses the merit of developing a Canadian siting bid. (author) 11 refs., 1 tab., 4 figs

  14. Experimental Study of Interfacial Friction in NaBH{sub 4} Solution in Microchannel Dehydrogenation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Seok Hyun; Hwang, Sueng Sik; Lee, Hee Joon [Kookmin Univ., Seoul (Korea, Republic of)

    2014-02-15

    Sodium borohydride (NaBH{sub 4}) is considered as a secure metal hydride for hydrogen storage and supply. In this study, the interfacial friction of two-phase flow in the dehydrogenation of aqueous NaBH{sub 4} solution in a microchannel with a hydraulic diameter of 461 μm is investigated for designing a dehydrogenation chemical reactor flow passage. Because hydrogen gas is generated by the hydrolysis of NaBH{sub 4} in the presence of a ruthenium catalyst, two different flow phases (aqueous NaBH{sub 4} solution and hydrogen gas) exist in the channel. For experimental studies, a microchannel was fabricated on a silicon wafer substrate, and 100-nm ruthenium catalyst was deposited on three sides of the channel surface. A bubbly flow pattern was observed. The experimental results indicate that the two-phase multiplier increases linearly with the void fraction, which depends on the initial concentration, reaction rate, and flow residence time.

  15. Experimental Study of Interfacial Friction in NaBH4 Solution in Microchannel Dehydrogenation Reactor

    International Nuclear Information System (INIS)

    Sodium borohydride (NaBH4) is considered as a secure metal hydride for hydrogen storage and supply. In this study, the interfacial friction of two-phase flow in the dehydrogenation of aqueous NaBH4 solution in a microchannel with a hydraulic diameter of 461 μm is investigated for designing a dehydrogenation chemical reactor flow passage. Because hydrogen gas is generated by the hydrolysis of NaBH4 in the presence of a ruthenium catalyst, two different flow phases (aqueous NaBH4 solution and hydrogen gas) exist in the channel. For experimental studies, a microchannel was fabricated on a silicon wafer substrate, and 100-nm ruthenium catalyst was deposited on three sides of the channel surface. A bubbly flow pattern was observed. The experimental results indicate that the two-phase multiplier increases linearly with the void fraction, which depends on the initial concentration, reaction rate, and flow residence time

  16. Computational and experimental studies of neutron spectra in the IGR reactor

    CERN Document Server

    Gorin, N V; Litvin, V I; Gajdajchuk, V A; Kazmin, Y M; Pakhnits, V A; Skivka, A S; Vasilev, A P; Pavshuk, V A; Rychev, A S

    2000-01-01

    The results of experiments made in order to determine spectral composition of neutrons in the IGR impulse graphite moderated reactor experimental channel at fuel temperature close to the room one are considered. The set of activation and fission detectors with half-life period more than 0.5 days is applied for the neutron spectrum measurements. The algorithm based on the directed divergence method is used for reconstruction of neutron energy spectra in energy range of 0.6 eV - 18 MeV. The results of calculational studies into the influence of impurities in structural materials on portion of thermal neutrons in he spectrum in the channel centre are discussed as well. The conclusion is made that the calculational results agree well with experimental data

  17. Turbulent precipitation of uranium oxalate in a vortex reactor - experimental study and modelling

    International Nuclear Information System (INIS)

    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)

  18. Experimental evaluation of feedforward control for the trajectory tracking of power in nuclear reactors

    International Nuclear Information System (INIS)

    This paper reports on an experimental comparison of feedforward control techniques for the trajectory-tracking of neutronic power was performed on the 5-MWt MIT Research Reactor. Included in the comparison were pure feedforward control in which the actuator signal is found solely by processing a demanded output through a system model, hybrid feedforward/feedback control in which the actuator signal is obtained by summing feedforward and feedback components, and period-generated control in which feedback is used to update the demand trajectory prior to its being processed through the system model for calculation of the actuator signal. This latter approach was found to be the most effective. In addition to the experimental results, discussions are given of both the rationale for model-based, feedforward control and the designs of the various controllers

  19. The Canadian initiative to bring the international thermonuclear experimental reactor to Canada

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) is the next step in fusion research. It is expected to be the last major experimental facility, before the construction of a prototype commercial reactor. The Engineering Design Activities (EDA) of ITER are being funded by the USA, Japan, the Russian Federation, and the European Union, with each of the major parties contributing about 25% of the cost. Canada participates as part of the European coalition. The EDA is due to be completed in 1998, and the major funding partners are preparing for the decision on the siting and construction of ITER. The Canadian Fusion Fuels Technology Project (CFFTP) formed a Canadian ITER Siting Task Group to study siting ITER in Canada. The study indicated that hosting ITER would provide significant benefits, both technological and economic, to Canada. We have also confirmed that there would be substantial benefits to the ITER Project. CFFTP then formed a Canadian ITER Siting Board, with representation from a broad range of stakeholders, to champion, 'Canada as Host'. This paper briefly outlines the ITER Project, and the benefits to both Canada and the Project of a Canadian site. With this as background, the paper discusses the international scene and assesses Canada's prospects of being chosen to host ITER. (author)

  20. Experimental study of flow and heat transfer in a rotating chemical vapor deposition reactor

    Science.gov (United States)

    Wong, Sun

    An experimental model was set up to study the rotating vertical impinging chemical vapor deposition reactor. Deposition occurs only when the system has enough thermal energy. Therefore, understanding the fluid characteristic and heat transfer of the system will provide a good basis to understand the full model. Growth rate and the uniformity of the film are the two most important factors in CVD process and it is depended on the flow and thermal characteristic within the system. Optimizing the operating parameters will result in better growth rate and uniformity. Operating parameters such as inflow velocity, inflow diameter and rotational speed are used to create different design simulations. Fluid velocities and various temperatures are recorded to see the effects of the different operating parameters. Velocities are recorded by using flow meter and hot wire anemometer. Temperatures are recorded by using various thermocouples and infrared thermometer. The result should provide a quantitative basis for the prediction, design and optimization of the system and process for design and fabrication of future CVD reactors. Further assessment of the system results will be discuss in detail such as effects of buoyancy and effects of rotation. The experimental study also coupled with a numerical study for further validation of both model. Comparisons between the two models are also presented.

  1. Experimental investigations on the migrational behaviour of silver in coated particle fuel for high-temperature reactors

    International Nuclear Information System (INIS)

    The migrational behaviour of silver in the coated particle fuel proposed for High-Temperature Reactors, is investigated experimentally. Data are described in the framework of the diffusion model. The diffussion coefficients are derived from the experimental data by a nonlinear least squares fit procedure. The experimental procedures and the theoretical calculations to analyse the data are described extensively. Arrhenius lines are presented for U(Th)-02, PyC and Sic. The silver release in advanced High-Temperature Reactors is prognosticated based on the measured data. (orig./HP)

  2. The computer science institute building of TU Brunswick University. Construction of an energy-efficient university building; Das Informatikzentrum der TU Braunschweig. Realisierung eines energieeffizienten Institutsgebaeudes

    Energy Technology Data Exchange (ETDEWEB)

    Rozynski, M.; Gerder, F. [Technische Univ. Braunschweig (Germany). Inst. fuer Gebaeude- und Solartechnik

    2003-07-01

    Saving of resources will be a key issue in future building construction. A new building projected on the campus of Brunswick University will have a power supply and ventilation concept that ensures low energy consumption. The project is carried out with funds provided by the Federal Minister of Economics and Technology (BMWi) in the context of the SolarBau funding concept. Construction of the building will be followed by an extensive monitoring programme that is to ensure its perfect function. [German] Eine wesentliche Zielsetzung zukuenftigen Bauens ist der sparsame Umgang mit Ressourcen. Im Rahmen eines integralen Planungsprozesses konnte fuer den Neubau des Informatikzentrums der TU Braunschweig ein Energie- und Lueftungskonzept realisiert werden, dass auf einen niedrigen Energieverbrauch zielt. Das Projekt wird im Rahmen des Foerderkonzeptes SolarBau durch das Bundesministerium fuer Wirtschaft und Technologie (BMWi) gefoerdert. Durch das anschliessende umfangreiche Monitoringprogramm wird derzeit die Funktionsfaehigkeit dieses Konzeptes ueberprueft. (orig.)

  3. Experimental direct digital control of the power plant A1 reactor based on a modern control theory approach

    International Nuclear Information System (INIS)

    The objective of the project was to accumulate technical experience with application of modern control theory in nuclear power by carrying out a case study of an experimental direct digital control at the A1 reactor about its nominal steady state. The research has proved that slightly modified method of solution of the linear stochastic regulator problem can be successfully applied in design of digital control system of nuclear power reactors

  4. First results with the experimental set-up at a Bugey reactor: neutrino oscillations, search of axions

    International Nuclear Information System (INIS)

    This work presents an experimental set-up at the Bugey PWR reactor to put into evidence neutrino oscillations. The first part describes a neutrino detector specially designed for the investigation of neutrino oscillations at two distances (13.50 m and 19 m) under the core of the reactor. Preliminary analysis are presented. The second part reports a search for axions, using the neutrino detector well-shielded volume. Created in competition with electro magnetic transitions, axion should be produced in abondance in the reactor core. This experiment excludes the existence of the axion of the standard model

  5. 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.

  6. Proceedings of the Workshop on Experimental and theoretical problems around actinides for future reactors

    International Nuclear Information System (INIS)

    Since the two last decades, in the framework of general researches on future reactors, strong efforts have been devoted to improve the quantity and quality of nuclear data. Indeed, in order to improve safety margins and fuel optimization, but also to develop new kind of reactors or fuel cycles, accurate nuclear data are mandatory. At the end of the twentieth century, nuclear data bases did not reach the required quality level to be used in future reactor simulations. Therefore, both experimentalists and theoreticians, in the framework of several European research programs (HINDAS, NUDATRA, ANDES, CHANDA...), have tried to make the situation better. New sets of precise data measurements concerning fission, capture, (n,xn),..., reaction cross sections for a large variety of nuclei have been initiated. From evaluation point of view, the JEFF project has also improved the quality of nuclear data bases for several nuclei. In parallel, on the theoretical side, progress has also been made concerning cross section modeling in a wide range of energy (eV to GeV). The goal was to provide theoretical models with a good predictive power to feed data bases where experimental data are still missing and where the measurement is too complex. In this context, for example, a new nuclear reaction code TALYS has been developed. Collaboration between experimentalists, theoreticians and evaluators are then of strong interest to make progress. The number of problems to be solved covers various fields of nuclear reactions such as fission, capture or inelastic scattering. In order to avoid too large an audience we have decided, as a first step, to focus on inelastic scattering on actinides. Experimentally, three main methods exist to measure the total inelastic cross section: activation, detection of the emitted neutrons and prompt-gamma spectroscopy. This last method is, nevertheless, dependent on theoretical models since it provides (n,xn γ) cross sections and not the total inelastic

  7. Oxidative coupling of methane in a fixed bed reactor over perovskite catalyst: A simulation study using experimental kinetic model

    Institute of Scientific and Technical Information of China (English)

    Nakisa Yaghobi; Mir Hamid Reza Ghoreishy

    2008-01-01

    The oxidative coupling of methane (OCM) to ethylene over a perovskite titanate catalyst in a fixed bed reactor was studied experimentally and numerically. The two-dimensional steady state model accounted for separate energy equations for the gas and solid phases coupled with an experimental kinetic model. A lumped kinetic model containing four main species CH4, O2, COx (CO2, CO), and C2 (C2H4 and C2H6) was used with a plug flow reactor model as well. The results from the model agreed with the experimental data. The model was used to analyze the influence of temperature and feed gas composition on the conversion and selectivity of the reactor performance. The analytical results indicate that the conversion decreases, whereas, C2 selectivity increases by increasing gas hourly space velocity (GHSV) and the methane conversion also decreases by increasing the methane to oxygen ratio.

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

    International Nuclear Information System (INIS)

    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

  9. Design study of an armor tile handling manipulator for the Fusion Experimental Reactor

    International Nuclear Information System (INIS)

    A conceptual design of the Fusion Experimental Reactor (FER), which is a D-T burning reactor following on JT-60 in Japan, has been developed by Japan Atomic Energy Research Institute (JAERI). In FER, a rail-mounted vehicle concept is planned to be adopted for in-vessel maintenance, such as maintenance of divertor plates and armor tiles. Advantages of this concept are the high stiffness of the rail as a base structure for maintenance and the high mobility of the vehicle along the rail. Twin armor tile handling manipulators installed on both sides of the vehicle have been designed. The respective manipulators for armor tile handling have 8 degrees of freedom in order to have access to any place of the first wall and to go through the horizontal port by operating manipulator joints. If the two types of manipulators for divertor plates and armor tiles are installed on the vehicle and the divertor handling manipulator carries a case filled with armor tiles, the replacement time of armor tiles will be reduced. In FER, moreover, maintenance of armor tiles, which is a scheduled maintenance, is planned to be carried out by the autonomous control using position sensors etc. In order to accumulate the data base for the development of the autonomous control of the manipulator in armor tile maintenance, the present paper describes basic mechanical characteristics (stress, deflection and natural frequency) of the armor tile handling manipulator calculated by static stress and dynamic eigenvalue analyses. (orig.)

  10. Audit of United States portion of the International Thermonuclear Experimental Reactor project

    International Nuclear Information System (INIS)

    Worldwide efforts in fusion energy research are designed to develop fusion power as a safe, environmentally sound, and economically competitive source of energy. The International Thermonuclear Experimental Reactor (ITER) project is a worldwide effort to demonstrate the scientific and technological feasibility of fusion power. The European Community, Japan, the Russian Federation, and the United States are collaborating on ITER, with each of the four parties expected to equally share costs and benefits. Shared costs for the current engineering design phase of the project are estimated at $1 billion in 1989 dollars, excluding certain management and support costs to be absorbed by each partner, with an early estimate of $6 billion, also in 1989 dollars, for construction of the reactor. Engineering design formally began in July 1992, and this phase is in its formative stages. The US had already spent about $100 million since 1987 on ITER conceptual design activities and other preparatory activities in advance of the engineering design phase. Because of its cost significance, the importance of ITER to the US fusion energy program, and the project's unique aspects which may provide a framework for future international endeavors, we initiated an audit of the ITER project. The purpose of the audit was to evaluate management controls over the US portion of the ITER project. Our objectives was to determine whether key front-end controls were in place to ensure that the project could be managed in an efficient and effective manner

  11. Requirements for US regulatory approval of the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) is the first fusion machine that will have sufficient decay heat and activation product inventory to pose potential nuclear safety concerns. As a result, nuclear safety and environmental issues will be much more important in the approval process for the design, siting, construction, and operation of ITER in the United States than previous fusion devices, such as the Tokamak Fusion Test Reactor. The purpose of this report is (a) to provide an overview of the regulatory approval process for a Department of Energy (DOE) nuclear facility; (b) to present the dose limits used by DOE to protect workers, the public, and the environment from the risks of exposure to radiation and hazardous materials; (c) to discuss some key nuclear safety-related issues that must be addressed early in the Engineering Design Activities (EDA) to obtain regulatory approval; and (d) to provide general guidelines to the ITER Joint Central Team (JCT) concerning the development of a regulatory framework for the ITER project

  12. Utilization of experimental integral data for the adjustment and uncertainty evaluation of reactor design quantities

    Energy Technology Data Exchange (ETDEWEB)

    Poenitz, W.P.; Collins, P.J.

    1988-01-01

    Biases and uncertainties of calculated reactor design quantities caused by errors and uncertainties of basic parameters, such as neutron cross sections, fission spectra parameters, and prompt and delayed neutron yields, are large, and in most cases, exceed reactor design requirements. Errors and uncertainties due to models and methods approximations contribute as well. An extensive data base, with presently /approximately/300 experimental integral values from 28 critical assemblies, has been assembled at Argonne National Laboratory in order to provide improvements and to investigate both sources of uncertainties. Generalized-least-squares fitting is being used. The available large data base permitted the investigation of the influence of specific input data, the constraints of the covariance information, the selection of parameters, and the reliability of the predictions. It is shown that reliable improvements of calculated quantities like enrichment, breeding ratio, sodium void, control rod worth, power distribution, and material worth can be made. Substantial reductions of the uncertainties of these quantities, which are caused by the uncertainties of the basic parameters, are obtained in most cases. The FFTF uranium-metal-core conversion is the first application of the present effort. 21 refs., 2 figs., 10 tabs.

  13. Design considerations for ITER [International Thermonuclear Experimental Reactor] plasma facing components

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) is a joint design and R ampersand D project involving the USA, the Soviet Union, Japan and the European Community. These international partners are working together on the design of a fusion tokamak reactor that will operate in the D-T ignition regime. This report compiles the contributions to ITER made by Sandia National Laboratories in the area of design and R ampersand D for plasma facing components, such as the first wall and divertor. The following topics are discussed: divertor fabrication issues, divertor thermal-hydraulic analysis, separatrix sweeping effects, divertor tile 2-D stress analysis, electromechanical disruption effects, runaway electron and intense energy deposition analyses, lifetime analysis and tritium retention in plasma facing materials. Material properties for pyrolytic graphite and beryllium are presented. Use of pyrolytic graphite as the plasma facing material allows for operation with thicker graphite armor at the design heat flux level of 10 MW/m2. The design of a divertor coated with plasma sprayed beryllium is presented as an attractive alternative to pyrolytic graphite armor tiles. Finally, the Sandia research and development plan for ITER is discussed. 82 figs

  14. Experimental fusion power reactor conceptual design study. Final report. Volume III

    International Nuclear Information System (INIS)

    This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

  15. 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.

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

    International Nuclear Information System (INIS)

    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

  17. ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  18. A full transport treatment of ITER [International Thermonuclear Experimental Reactor] burn control scenarios

    International Nuclear Information System (INIS)

    Reference operating scenarios for the physics phase of the International Thermonuclear Experimental Reactor (ITER) rely on low temperature, high density operation for good divertor performance. Operating points in this region are usually thermally unstable and active burn control is required to maintain operation at the selected operating point. We present a series of transport simulations of ITER burn control scenarios with modulated neutral beam heating, using the 1-1/2D code WHIST. Our results indicate that control is possible, except at very high densities (left-angle ne right-angle ≥ 1.80x1020 m-3) where off-axis heating due to the poor beam penetration makes control difficult, especially for negative perturbations 15 refs., 20 figs

  19. Theoretical and experimental study of collectrons for epithermal neutron flux in reactors

    International Nuclear Information System (INIS)

    A theoretical study of nuclear reactions and electric charge displacements arising in sensitivity to thermal and epithermal neutrons in collectrons allowed a computer code conception. Collectrons in Rhodium, Silver, Cobalt, Hafnium, Erbium, Gadolinium and Holmium have been tested in different radiation fields given by neutron or gamma filters irradiated in different places of Melusine and Siloe reactors. Some emitters were covered with different steel, nickel or zircaloy thicknesses. Theoretical and experimental results are consistent; that validate the computer code and show possibilities and necessity of covering collectron emitters to reduce or cancel the gamma sensitivity and to improve response instantaneity. A selective measurement of epithermal neutron flux can by this way, made by associating two types of collectrons

  20. A method of simulating voids in experimental studies of boiling water reactors

    International Nuclear Information System (INIS)

    The coolant density in boiling water reactors may vary from 3 at pressures up to 1000 p.s.i. In order to study the effect of reduced water density on reactivity in unpressurized experimental systems, the effective water density is reduced by packing small beads of highly expanded polystyrene into the fuel clusters and flooding the interstices with water. Coolant densities of from 0.4 to 0.6 gm/cm3 may be produced with the introduction of only about 0.4 gm/cm3 of non-hydrogeneous material. This memorandum describes the production, properties and handling of polystyrene beads and the tests carried out to establish the validity of the technique. (author)

  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. Experimental Research for Refractories Used in High-Turbulent-Mixer (HTM)-Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The high turbulent mixer (HTM) process is benefit for improving the quality of iron and steel, but the refractory of the reactor is easy to be eroded by slag and liquid metal at high temperature. Especially for the violent mixing in HTM, refractory can also be impacted by molten metal (iron or steel), so it is very important to study and find out new refractory to meet the demand of HTM. Suitable refractory not only can stand the eroding of slag and liquid metal, but also can reduce the loss of electromagnetic energy. According to the experimental results, the influence of different refractory on electromagnetic force is unconspicuous, the refractory with Al2O3-base is better than that with MgO-base for standing the erosion by slag and iron. Al2O3-base refractory is more suitable for HTM process.

  3. Development of 100 GHz band high power gyrotron for fusion experimental reactor

    International Nuclear Information System (INIS)

    In JAERI, 1MW gyrotrons of 170GHz and 110GHz are under development for ITER (International Thermonuclear Experimental Reactor) and JT-60U, respectively. Both gyrotrons have a depressed collector for an efficiency improvement and a low loss synthetic diamond window that enables Gaussian beam output over 1MW. Three 110GHz gyrotrons are used on an electron cyclotron heating and current drive(ECH/ECCD) system on JT-60U, in which the output power of ∼0.8MW/3sec was generated from each gyrotron. As for 170GHz, output power of 1.2MW with electron beam of 85kV/49A was obtained on a short pulse gyrotron. The efficiency of ∼57% was attained at 1.1MW with the depressed collector. Based on these results, the 1MW 170GHz gyrotron for long pulse operation was fabricated. (author)

  4. NAIADQ, a computer program for calculating reactivity transients in low power experimental water reactors

    International Nuclear Information System (INIS)

    The computer code NAIADQ is designed to simulate the course and consequences of non-destructive reactivity accidents in low power, experimental, water-cooled reactor cores fuelled with metal plate elements. It is a coupled neutron kinetics-hydrodynamics-heat transfer code which uses point kinetics and one-dimensional thermohydraulic equations. Nucleate boiling, which occurs at the fuel surface during transients, is modelled by the growth of a superheated layer of water in which vapour is generated at a non-equilibrium rate. It is assumed that this vapour is formed at its saturation temperature and that it mixes homogeneously with the water in this layer. The code is written in FORTRAN IV and has been programmed to run as a catalogued procedure on an IBM operating system such as MVT or MVS, with facility for the inclusion of user routines

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

    International Nuclear Information System (INIS)

    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

  6. Development of Nb3Sn strands for the International Thermonuclear Experimental Reactor (ITER) in Japan

    International Nuclear Information System (INIS)

    Development programs focusing on Nb3Sn strands for the International Thermonuclear Experimental Reactor (ITER) are being performed by each party of the ITER group. The object of development is not only to achieve high superconductor performance such as high current density and low hysteresis loss but also enable reliable low-cost mass production, one way of which is to reduce strand breaking during the manufacturing process, and long-length strand fabrication. In Japan, the Japan Atomic Energy Research Institute (JAERI) is involved in this program backed by strong collaboration with many domestic industries. Totally, 11.1 tons (2,400 km-length) of strands that meet ITER specifications have been fabricated successfully. Through this development work, the Nb3Sn strand mass-production technique in Japan has been substantially improved. (author)

  7. Safety analysis of fuel circulating system for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study and safety analysis on the fuel circulating system (FCS) for a Tokamak experimental fusion reactor have been carried out, in tritium containment and safety. A three-stage containment system of tritium fuel is settled to control tritium release to the environment as low as possible. The system safety is examined by fault trees, failure effect and related analyses. FCS components and piping are designed to minimize leakage into the secondary containment which encloses most of the primary containment. The secondary containment is of glovebox type with an inert gas atmosphere, provided Glovebox Atmosphere Purification System of flow rate 70 m3/min. The tertiary containment system consists of the rooms enclosing secondary system and tritium-bearing gas processing systems, a 3 m3/min Atmosphere Cleanup System and a 150 m3/min Emergency Containment System. (author)

  8. RIA and LOCA simulating tests on experimental fuel elements in TRIGA MT reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Full text: One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident condition. A total of 39 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 100 test fuel elements have been irradiated in TRIGA SS MTR in different power history conditions. LOCA simulating tests are planned to be performed in C2 LOCA tests capsule and in Loop A of TRIGA SS MTR of INR Pitesti. The LOCA tests in capsule C2 are instrumented to measure fuel, sheath and coolant temperature, internal element and coolant pressure during the entire irradiation period. In the second phase of the experiment the C2 capsule will be connected to the sweep gas system with the on-line gamma ray spectrometer included. RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. This paper

  9. 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.

  10. Characterization and Application of the Thermal Neutron Radiography Beam in the Egyptian Second Experimental and Training Research Reactor (ETRR-2)

    OpenAIRE

    M. A. Abou Mandour; R. M. Megahid; Hassan, M.H.; T. M. Abd El Salam

    2007-01-01

    The Experimental, Training, Research Reactor (ETRR-2) is an open-pool multipurpose reactor (MPR) with a core power of 22 MWth cooled and moderated by light water and reflected with beryllium. It has four neutron beams and a thermal column as the main experimental devices. The neutron radiography facility unit utilizes one of the radial beam tubes. The track-etch technique using nitrocellulose films and converter screen is applied. In this work, the radial neutron beam for the thermal neutron ...

  11. An experimental and modeling investigation of particle production by spray pyrolysis using a laminar flow aerosol reactor

    International Nuclear Information System (INIS)

    The influence of operating parameters on the morphology of particles prepared by spray pyrolysis was investigated using a temperature-graded laminar flow aerosol reactor. Experimentally, zirconia particles were prepared by spray pyrolysis using an aqueous solution of zirconyl hydroxide chloride. Hollow particles were formed if the reactor temperature was high, the temperature gradient was too large, the flow rate of carrier gas was high, and the initial solute concentration was low. A numerical simulation of the pyrolysis process was developed using a combination of two previous models. The simulation results compared well with the experimental results. (c) 2000 Materials Research Society

  12. 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.

  13. A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

    International Nuclear Information System (INIS)

    A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

  14. Engineering solutions for components facing the plasma in experimental fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Farfaletti-Casali, F.

    1986-07-01

    An analysis is made of the engineering problems related to the structures facing the plasma in experimental tokamak-type reactors. Attention is focused on the so-called ''current first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system. The design of a first wall, developed at the JRC-Ispra for INTOR/NET and based on the idea of conceiving it as one of the sides, of a box which envelopes a blanket segment, is described. The progress in the structural analysis of the first wall box under operating and abnormal (plasma disruption) conditions is presented and discussed. The design of the collector plates of the single-null divertor of INTOR/NET, as developed at the JRC-Ispra, is described. This design is based on a W-Re protective layer and a water-cooled heat sink, including cooling channels iun Cu-alloys and a Cu-matrix for bonding. The results of the elastic and elasto-plastic evaluations are discussed, together with a layout of the experimental activity in progress. It is concluded that, even if the uncertainties related to the plasma-wall interaction are still relevant, the engineering solutions identified look manageable, although they require a large research and development effort.

  15. Reactor physics calculations and their experimental validation for conversion and upgrading of a typical swimming pool type research reactor

    International Nuclear Information System (INIS)

    Detailed neutronic analysis of a typical swimming pool type research reactor, Pakistan Research Reactor-1 (PARR-1), was carried out for conversion of its core from 93% highly enriched uranium to 20% low enriched uranium fuel with power upgrading from 5 to 10 MW. Standard computer codes WIMS-D/4 and CITATION were employed to calculate core excess reactivity, power defect, reactivity effect of xenon and samarium, reactivity worth of fuel element, worth of control rods, shutdown margin, reactivity feedback coefficients, neutron flux and power peaking factors. A series of low and high power tests were performed on the newly converted core to determine its performance. A comparison between the calculated and measured results is presented in this article. The agreement is generally good

  16. Experimental investigation of fluidized-bed reactor performance for oxidative coupling of methane

    Institute of Scientific and Technical Information of China (English)

    S.Ja(s)o; R.Schom(a)cker; G.Wozny; S.Sadjadi; H.R.Godini; U.Simon; S.Arndt; O.G(o)rke; A.Berthold; H.Arellano-Garcia; H.Schubert

    2012-01-01

    Performance of the oxidative coupling of methane in fluidized-bed reactor was experimentally investigated using Mn-Na2WO4/SiO2,La2O3/CaO and La2O3-SrO/CaO catalysts.These catalysts were found to be stable,especially Mn-Na2WO4/SiO2 catalyst.The effect of sodium content of this catalyst was analyzed and the challenge of catalyst agglomeration was addressed using proper catalyst composition of 2%Mn-2.2%Na2WO4/SiO2.For other two catalysts,the effect of Lanthanum-Strontium content was analyzed and 10%La2O3-20%SrO/CaO catalyst was found to provide higher ethylene yield than La2O3/CaO catalyst.Furthermore,the effect of operating parameters such as temperature and methane to oxygen ratio were also reviewed.The highest ethylene and ethane (C2) yield was achieved with the lowest methane to oxygen ratio around 2.40.5% selectivity to ethylene and ethane and 41% methane conversion were achieved over La2O3-SrO/CaO catalyst while over Mn-Na2WO4/SiO2 catalyst,40% and 48% were recorded,respectively.Moreover,the consecutive effects of nitrogen dilution,ethylene to ethane production ratio and other performance indicators on the down-stream process units were qualitatively discussed and Mn-Na2WO4/SiO2 catalyst showed a better performance in the reactor and process scale analysis.

  17. Conceptual design and technology development of containment structure in Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    A conceptual design of FER (Fusion Experimental Reactor) containment structure and its associated R and D activities, conducted from '89 to '90, are described. The FER containment structure system which mainly consists of a vacuum vessel, shielding structures, in-vessel replaceable components, ports, a cooling pipe system, has been developed to fullfil the required function. As an initial stage of R and D activities, the elemental technologies common to a tokamak reactor have been developed. Among them, a locking mechanism for supporting in-vessel replaceable components and a technique for insulation/conduction are described. For the locking mechanism, a caulking cotter driven by hydraulic pressure has been employed. Three kinds of hydraulic driving mechanism have been manufactured by trial: a 'piston jack' type, a 'bellows' type and a 'flexible tube' type. In the latter type, the stroke is obtained by changing the cross section of the flexible tube from a flat racetrack shape to a fat shape by hydraulic pressure. As the result of preliminary performance test, the shape of 'flexible tube' has been found to be improved. For the insulation coating, Al2O3 has been selected as the material and a plasma spray method has been applied as the coating procedure. For the conduction coating, Cr3C2 has been selected as the material and JET-KOTE method has been applied as the coating procedure. Both methods have been successfully developed and have been confirmed to be applicable the actual machine. A one fifth scale model has been fabricated in order to verify the design feasibility, mainly geometrical consistency. Then some design modifications were found to be needed for some of the components based on the manufacturing experience. (author)

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

    International Nuclear Information System (INIS)

    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.

  19. 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.

  20. Experimental and analytical investigations of primary coolant pump coastdown phenomena for the Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alatrash, Yazan [Advanced Nuclear Engineering System Department, Korea University of Science and Technology (UST), 217 Gajeong-ro Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Kang, Han-ok; Yoon, Hyun-gi; Seo, Kyoungwoo; Chi, Dae-Young [Korea Atomic Energy Institute (KAERI), 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Yoon, Juhyeon, E-mail: yoonj@kaeri.re.kr [Korea Atomic Energy Institute (KAERI), 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), Daejeon (Korea, Republic of)

    2015-05-15

    Highlights: • Core flow coastdown phenomena of a research reactor are investigated experimentally. • The experimental dataset is well predicted by a simulation software package, MMS. • The validity and consistency of the experimental dataset are confirmed. • The designed coastdown half time is confirmed to be well above the design requirement. - Abstract: Many low-power research reactors including the Jordan Research and Training Reactor (JRTR) are designed to have a downward core flow during a normal operation mode for many convenient operating features. This design feature requires maintaining the downward core flow for a short period of time right after a loss of off-site power (LOOP) accident to guarantee nuclear fuel integrity. In the JRTR, a big flywheel is installed on a primary cooling system (PCS) pump shaft to passively provide the inertial downward core flow at an initial stage of the LOOP accident. The inertial pumping capability during the coastdown period is experimentally investigated to confirm whether the coastdown half time requirement given by safety analyses is being satisfied. The validity and consistency of the experimental dataset are evaluated using a simulation software package, modular modeling system (MMS). In the MMS simulation model, all of the design data that affect the pump coastdown behavior are reflected. The experimental dataset is well predicted by the MMS model, and is confirmed to be valid and consistent. The designed coastdown half time is confirmed to be well above the value required by safety analysis results. (wwwyoon@gmail.com)

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

    OpenAIRE

    Noble Brooklyn; Choe Dong-Ok; Jevremovic Tatjana

    2012-01-01

    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 f...

  2. Thermal Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Water: Performance and Stability

    Science.gov (United States)

    Lisowski, Darius D.

    This experimental study investigated the thermal hydraulic behavior and boiling mechanisms present in a scaled reactor cavity cooling system (RCCS). The experimental facility reflects a ¼ scale model of one conceptual design for decay heat removal in advanced GenIV nuclear reactors. Radiant heaters supply up to 25 kW/m2 onto a three parallel riser tube and cooling panel test section assembly, representative of a 5° sector model of the full scale concept. Derived similarity relations have preserved the thermal hydraulic flow patterns and integral system response, ensuring relevant data and similarity among scales. Attention will first be given to the characterization of design features, form and heat losses, nominal behavior, repeatability, and data uncertainty. Then, tests performed in single-phase have evaluated the steady-state behavior. Following, the transition to saturation and subsequent boiling allowed investigations onto four parametric effects at two-phase flow and will be the primary focus area of remaining analysis. Baseline conditions at two-phase flow were defined by 15.19 kW of heated power and 80% coolant inventory, and resulted in semi-periodic system oscillations by the mechanism of hydrostatic head fluctuations. Void generation was the result of adiabatic expansion of the fluid due to a reduction in hydrostatic head pressure, a phenomena similar to flashing. At higher powers of 17.84 and 20.49 kW, this effect was augmented, creating large flow excursions that followed a smooth and sinusoidal shaped path. Stabilization can occur if the steam outflow condition incorporates a nominal restriction, as it will serve to buffer the short time scale excursions of the gas space pressure and dampen oscillations. The influences of an inlet restriction, imposed by an orifice plate, introduced subcooling boiling within the heated core and resulted in chaotic interactions among the parallel risers. The penultimate parametric examined effects of boil-off and

  3. Experimental study of thermohydraulic processes and gas distribution in a model of the containment shell of the AST-500 reactor

    International Nuclear Information System (INIS)

    Experiments were made on a setup consisting of a large-scale twin-assembly model of the primary circuit of an integral reactor and of a model of a containment shell which is a means for confining the outflow of coolant from the reactor. The large-scale model of an AST-500 reactor has vertical dimensions close to the actual dimensions and similar coefficients of hydraulic resistance and volume ratios of the principal elements the circuit with natural circulation. The model of the containment shell is a vertical cylindrical vessel with a size of 426 x 12 mm, a height of 9.78 m, and a volume of 1.24 m3. The volume scale of the reactor model and of the model of the containment shell is 1:170. The elements of the latter model are made from steel 20. The models of the reactor and of the containment shell are joined through two pipelines with a size of 57 x 3.5 mm and shut-off valves with a diameter of 50 mm mounted thereon. A total of 70 experiments were made to simulate leakage of the primary circuit of the integrated reactor and the outflow of coolant into the containment shell. The authors have provided detailed information on the large-scale model, have described the experimental conditions, and have reported on the main results of their study of the development of an accident involving the loss of coolant in the reactor-containment shell system. The present article reports on a study of the thermohydraulic processes and the gas distribution in the containment shell. Since the designs of the model and of the actual containment shell of the AST-500 reactor are not identical, the authors assume that the results reported can be used in appropriate computer programs describing the processes which occur in containment vessels of atomic power stations (containment shells, protective shells, sealed assemblies)

  4. 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.

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

    Science.gov (United States)

    Destouches, Christophe

    2016-03-01

    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.

  6. Experimental simulation study on hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    The hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor is studied through a 1:2.33 test model. The design and other feature of the test model is described. The experimental results show that the flow resistance coefficient of the heat exchanger becomes self-simulation when Reynolds number is greater than 5000. The value of flow resistance coefficient at self-simulation condition and the distribution of pressure drop in the heat exchanger are given through experiment. The option design to reduce flow resistance is proposed. The designed and experimental value for the flow resistance coefficient are in good agreement. The variation of system parameters during flow excursion was described. The experimental results are of great significant for the final design of the main heat exchanger of Daqing 200 MW nuclear heating reactor. (2 refs., 5 figs., 1 tab.)

  7. Modelling activities of experimental facilities related to advanced reactors. Considerations on 1D/3D issues

    International Nuclear Information System (INIS)

    The state of art of modelling activities related to integral experimental facilities of advanced passive reactors show to date important open items. The main advantage of using 1D plant codes is the capability of simulating the full interaction between components traditionally correctly modelled (condensers, heat exchangers, pipes and vessels) and other components for which codes are not 100% suitable (pools and containments). Polytechnical University of Catalonia (UPC) and Polytechnical University of Valencia (UPV) cooperated with other European research organizations in the 'Technology Enhancement for Passive Safety Systems' (TEPSS) project, within the European Fourth Framework Programme. It was a task of both Universities to supply analytical support of PANDA tests. The paper deals with the 1D/3D discussion in the framework of modelling activities related to integral passive facilities like PANDA. It starts choosing reference tests among those corresponding to our participation in TEPSS project. The discrepancies observed in a 1D simulation of the selected tests will be shown and analyzed. An evaluation of how the 3D version can lead to a better agreement with data will be included. Disadvantages of 3D codes will be shown too. Combining the use of different codes, and considering analyst criteria, will make possible to establish suitable recommendations from both engineering and scientific point of view. (author)

  8. 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.

  9. Sensitivity analysis of neutronics calculations in the preliminary design of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sensitivity of principal neutronics characteristic quantities for the neutron cross sections of JAERI Experimental Fusion Reactor (JXFR) has been studied by means of sensitivity analysis method based on linear perturbation theory. The same study was made previously. After publication of the previous results, however, the SWANLAKE code used to calculate sensitivities was found to include error derived during its conversion process. The study was thus repeated with corrected SWANLAKE. The quantities studied are calculational results for the first preliminary design of JXFR such as the (n, p) reaction rates of 58Ni and 54Fe in the outer part of superconducting toroidal field coil (TFC), the copper atomic displacement rate in the inner part of TFC and the tritium production rate in the outer blanket. Though the calculational results do not contradict essentially the results in the former study, the newly calculated sensivitities were found to be more or less different from the previous ones. Therefore, the results and discussion of analysis given in this report are revised, with the values corrected. The errors of the (n, p) reaction rates and the copper displacement rate due to the uncertainties of cross sections were estimated to be about 50 - 70% and 25 - 65%, respectively, taking into account the direct sensitivity of (n, p) reaction cross sections in the former. (author)

  10. Report on Task 2 in the special working group of the International Thermonuclear Experimental Reactor (ITER)

    Energy Technology Data Exchange (ETDEWEB)

    Kishimoto, Hiroshi [Japan Atomic Energy Research Inst., Tokyo (Japan)

    1999-10-01

    Under international cooperation of four parties of U.S.A., EU, Russia, and Japan, the Engineering Design Activity (EDA) on the International Thermonuclear Experimental Reactor (ITER) began in June, 1992, the actions were finished their initial six years programs in 1998, but were not transferred to its construction. Then, an investigation on design option to intend to reduce its construction cost to it's a half was conducted by three years elongation of EDA, for which a Special Working Group (SWG) was established at the 13th ITER boards of directors. The work of this Task 2 was to evaluate some impacts to the Nuclear Fusion Development Program. The report of the Task 2 consists of five chapters, where general nuclear fusion development were discussed at a center of Tokamak. In special, on Chapter 4: Comparison with the other routes, for a modular route claimed by U.S.A., some earnest discussions were conducted by U.S.A. and other three parties, to be anxious to for U.S.A. to suggest to withdraw this work. In this report, four points of Japanese claims on security of flexibility capable of conducting advanced Tokamak research in ITER, safety proof due to positive introduction and test of low activation materials to ITER, and so forth, were all reflected. In the SWG, an opinion that the nuclear fusion program in the world was prepared to start to ITER was agreed by all members. (G.K.)

  11. Report on Task 2 in the special working group of the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Under international cooperation of four parties of U.S.A., EU, Russia, and Japan, the Engineering Design Activity (EDA) on the International Thermonuclear Experimental Reactor (ITER) began in June, 1992, the actions were finished their initial six years programs in 1998, but were not transferred to its construction. Then, an investigation on design option to intend to reduce its construction cost to it's a half was conducted by three years elongation of EDA, for which a Special Working Group (SWG) was established at the 13th ITER boards of directors. The work of this Task 2 was to evaluate some impacts to the Nuclear Fusion Development Program. The report of the Task 2 consists of five chapters, where general nuclear fusion development were discussed at a center of Tokamak. In special, on Chapter 4: Comparison with the other routes, for a modular route claimed by U.S.A., some earnest discussions were conducted by U.S.A. and other three parties, to be anxious to for U.S.A. to suggest to withdraw this work. In this report, four points of Japanese claims on security of flexibility capable of conducting advanced Tokamak research in ITER, safety proof due to positive introduction and test of low activation materials to ITER, and so forth, were all reflected. In the SWG, an opinion that the nuclear fusion program in the world was prepared to start to ITER was agreed by all members. (G.K.)

  12. 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.

  13. Bulk-bronzied graphites for plasma-facing components in ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 NbSn3 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

  15. Development of remote dimensional measurement system for first wall of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    A remote system has been developed for detection and dimensional measurement of the eroded defect of protection tiles attached to first walls of the Fusion Experimental Reactor (FER). The test system is composed of a vision sensor and a robot to move it. The sensor uses a laser slit projector and an TV camera, and measures the tile surface three-dimensionally and untouchly based on the triangulation method. In order to protect sensitive components such as a semiconductor laser source and a CCD camera from high radiation field in the vacuum vessel, a sensor head to be inserted in the vacuum vessel is equipped with only the projector and a zoom lens, which are connected to the laser source and the camera by a light guide and an imagefiber respectively. A defect, which has been detected from measurement in wide area of the tiles with the zoom lens of wide position, is measured with the zoom lens of telescopic position by execution of the robot motion planned autonomously. As a result of measurement tests, it has been proved that the system can detect the defect with sufficient precision and is, therefore, practicable. (author)

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

    International Nuclear Information System (INIS)

    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

  17. Oak Ridge Tokamak Experimental Power Reactor study, 1976. Part I. EPR summary

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, M.; McAlees, D.G.; Shannon, T.E.; Flanagan, C.A.; Lue, J.W.

    1977-04-01

    The study of the Tokamak Experimental Power Reactor (EPR) has been performed by the Oak Ridge National Laboratory (ORNL) for the Energy Research and Development Administration Division of Magnetic Fusion Energy (ERDA-DMFE) over a two-year period. This second year's work builds integrally upon the earlier efforts. Therefore, the following format has been adopted for this report. Part 1, entitled ''EPR Summary,'' contains a brief discussion of all facets of the work done this year. The technical areas discussed in Part 1 include the Reference Design itself; the underlying plasma engineering base; the supporting technical evaluations in the areas of magnet systems, nuclear engineering, and general engineering; and the major research, development, and demonstration needs that have been identified. Each of these areas is then discussed fully in separate technical memoranda, which have been issued as Parts 2-6. Where appropriate, the essential elements of previous interim reports have been included in order to provide a self-consistent, readable report. These earlier reports, Basic Considerations and Initiation of Studies (ORNL/TM-4853), EPR Scoping Study (ORNL/TM-5038), and EPR Reference Design (ORNL/TM-5042), can be consulted for a more complete discussion of the entire study.

  18. Recommendations for a cryogenic system for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Soviet Union, and the United States. ITER will be a large machine requiring up to 100 kW of refrigeration at 4.5 K to cool its superconducting magnets. Unlike earlier fusion experiments, the ITER cryogenic system must handle pulse loads constituting a large percentage of the total load. These come from neutron heating during a fusion burn and from ac losses during ramping of current in the PF (poloidal field) coils. This paper presents a conceptual design for a cryogenic system that meets ITER requirements. It describes a system with the following features: (a) only time-proven components are used; (b) the system obtains a high efficiency without use of cold pumps or other developmental components; (c) high reliability is achieved by paralleling compressors and expanders and by using adequate isolation valving; (d) the problem of load fluctuations is solved by a simple load leveling device; (e) the cryogenic system can be housed in a separate building located at a considerable distance from the ITER core, if desired. The paper also summarizes physical plant size, cost estimates, and means of handling vented helium during a magnet quench. 4 refs., 4 figs., 3 tabs

  19. Review of the conceptual design of a Doublet fusion experimental power reactor

    International Nuclear Information System (INIS)

    The results of a two-year, conceptual design study of a fusion experimental power reactor (EPR) are presented. For this study, the primary objectives of the EPR are to obtain plasma ignition conditions and produce net electrical power. The design features a Doublet plasma configuration with a major radius of 4.5 m. The average plasma beta is 10 percent which yields a thermonuclear power level of 410 MW during a 105-sec burn period. With a duty factor of 0.84, the gross electrical output is 124 MW(e) while the net output is 37 MW(e). The design features a 25-cm-thick, helium-cooled, modular, stainless-steel blanket with a 1-cm-thick, silicon carbide first wall. Sufficient shielding is provided to permit contact maintenance outside the shield envelope within 24 hr after shutdown. An overall plant concept has been developed including a superheated steam cycle power conversion system. Preliminary cost estimates and construction schedules have also been developed. 3 refs

  20. 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.

  1. Cross-section sensitivity analyses for a Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Simmons, E.L.; Gerstl, S.A.W.; Dudziak, D.J.

    1977-09-01

    The objectives of this report were (1) to determine the sensitivity of neutronic responses in the preliminary design of the Tokamak Experimental Power Reactor by Argonne National Laboratory, and (2) to develop the use of a neutron-gamma coupled cross-section set in the calculation of cross-section sensitivity analysis. Response functions such as neutron plus gamma kerma, Mylar dose, copper transmutation, copper dpa, and activation of the toroidal field coil dewar were investigated. Calculations revealed that the responses were most sensitive to the high-energy group cross sections of iron in the innermost regions containing stainless steel. For example, both the neutron heating of the toroidal field coil and the activation of the toroidal field coil dewar show an integral sensitivity of about -5 with respect to the iron total cross sections. Major contributors are the scattering cross sections of iron, with -2.7 and -4.4 for neutron heating and activation, respectively. The effects of changes in gamma cross sections were generally an order of 10 lower.

  2. 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

  3. 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.

  4. Design study of plant system for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

  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. Simulating Experimental Investigation on the Safety of Nuclear Heating Reactor in Loss—of —Coolant Accidents

    Institute of Scientific and Technical Information of China (English)

    ZhanjieXu

    1996-01-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology(INET) of Tisinghuan University of CHina in 1989,Its main loop is a thermal-hydraulic system with natural circulation.This paper studies the safety of NHR under the condition of loss-of -coolant accidents(LOCAs) by means of simulant experiments.First,the Background and necessity of the experiments are presented.then the experimental system,including the thermal-hydraulic system and the data collection system,and similarity criteria are introduced.Up to now ,the discharge experiments with the residual heating power(20% rated heating power)have been carried out on the experimental system,The system prameters including circulation flow rate,system pressure,system temperature,void fraction,discharge mass and so on have been recorded and analyzed.Based on the results of the experiments,the conclusionas are shown as follos:on the whole,the reactor is safe under the condition of LOCAs,but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  7. An experimental study on in-vessel retention strategy by external reactor vessel cooling with liquid metal

    International Nuclear Information System (INIS)

    The present work ultimately aims to develop the IVR-ERVC (In-Vessel Retention through External Reactor Vessel Cooling) system with enough thermal margin adopting liquid metal coolant as the severe accident mitigation system even for high power reactor. For the purpose, the conceptual design of IVR-ERVC with liquid metal is evaluated by performing an experimental campaign for a scaled facility. The specific geometry was devised to contain the liquid metal coolant facing with water through the container wall. Through this system, the heat transfer area is enlarged up to 2 times compared to the original area of the reactor vessel. This effect is also named as 'liquid metal fin' in the current study. Improved heat transfer or reduced heat flux including large drop of focusing effect was confirmed by experimental results for a small-scaled facility to simulate the boiling phenomena under IVR-ERVC condition. It was found that significant reduction of focusing effect by liquid metal and extended surface area guarantee enough margin of successful IVR-ERVC without CHF issue even for large-sized power reactors. (author)

  8. Demonstration test of the holding stability of the self actuated shutdown system in the experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Self actuated shutdown system (SASS) with a Curie point electromagnet (CPEM) has been developed for use in a large scale fast breeder reactor (FBR) in order to establish the passive shutdown capability against anticipated transient without scram (ATWS) events. The basic characteristics of SASS have already been investigated by various out-of-pile tests for material elements. As the final stage of the development, the stability of SASS needs to be confirmed under the actual reactor-operational environment with high temperature, high neutron flux, and flowing sodium in order to ensure the high plant availability factor. For this purpose, the demonstration test of holding stability using the reduced-scale experimental equipment of SASS was conducted in the 1st and 2nd operational cycles of the experimental fast reactor JOYO MK-III. As a result of this study, the rod-holding stability and the rod-recovering functions of the driving system to re-connect and pull out the separated control rod were fully confirmed. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. (author)

  9. Experimental and analytical study on thermal hydraulics in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Akimoto, Hajime; Araya, Fumimasa; Ohnuki, Akira; Yoshida, Hiroyuki; Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    Study and development of reduced-moderation spectrum water reactor proceeds as a option of the future type reactor in Japan Atomic Energy Research Institute (JAERI). The reduced-moderation spectrum in which a neutron has higher energy than the conventional water reactors is achieved by decreasing moderator-to-fuel ratio in the lattice core of the reactor. Conversion ratio in the reduced-moderation water reactor can be more than 1.0. High burnup and long term cycle operation of the reactor are expected. A type of heavy water cooled PWR and three types of BWR are discussed as follows; For the PWR, (1) critical heat flux experiments in hexagonal tight lattice core, (2) evaluation of cooling limit at a nominal power operation, and (3) analysis of rewetting cooling behavior at loss of coolant accident following with large scale pipe rupture. For the BWR, analyses of cooling limit at a nominal power operation of, (1) no blanket BWR, (2) long term cycle operation BWR, and (3) high conversion ratio BWR. The experiments and the analyses proved that the basic thermal hydraulic characteristics of these reduced-moderation water reactors satisfy the essential points of the safety requirements. (Suetake, M.)

  10. Monte Carlo simulation analysis of integral data measured in the SCK-CEN/ENEA experimental campaign on the TAPIRO fast reactor. Experimental and calculated data comparison

    Energy Technology Data Exchange (ETDEWEB)

    Burgio, N., E-mail: nunzio.burgio@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), C.R. Casaccia Via Anguillarese 301, 00123 Rome (Italy); Cretara, L., E-mail: luca.cretara@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Frullini, M., E-mail: massimo.frullini@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Gandini, A., E-mail: augusto.gandini@uniroma1.it [DIAEE – Sapienza University of Rome, Corso Vittorio Emanuele II 244, 00186 Rome (Italy); Peluso, V., E-mail: vincenzogiuseppe.peluso@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), Via Martiri di monte Sole 4, 40129 Bologna (Italy); Santagata, A., E-mail: alfonso.santagata@enea.it [ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development), C.R. Casaccia, Via Anguillarese 301, 00123 Rome (Italy)

    2014-07-01

    Highlights: • We develop a MCNPX model of the TAPIRO fast research reactor. • The model has been tested against the result of a late experimental champaign finding on overall agreement. • The source of uncertainties in the nuclear data and in the model assumptions has been discussed. • The model is sufficiently accurate to design irradiation experiment in support to R and D activities on LFR and ADS systems. - Abstract: After Fukushima events, the Italian nuclear program has been redefined leaving space only to activities related to Generation IV nuclear systems. Accordingly with this renewed national scenario, TAPIRO fast reactor facility is gaining a relatively major strategic role. A program is in fact being proposed to host in TAPIRO benchmark experimental activities related to the development of Lead fast reactor and Accelerator Driven Systems. A first step of this program would consist on the validation of neutronic codes, cross section data and reactor models to be adopted for its analysis. Along this line in this work the results of a simulation study has been made relevant to the measurements performed in the SCK-CEN/ENEA experimental campaign carried out in the 1980–1986 period. The calculations have been made using the Monte Carlo MCNPX 2.7.0 Code. In this article the main results are presented and discussed, with particular emphasis on the uncertainties, relevant both to nuclear data and the model layout. The results of this simulation study indicate in particular that TAPIRO's MCNPX model is adequate for the optimization of set-ups of perspective neutron irradiation experiments, this allowing cuts in costs and development time.

  11. Experimental fast reactor Joyo emergency operation act in Northeastern Japan Earthquake 3.11

    International Nuclear Information System (INIS)

    Experimental fast reactor 'Joyo' under facility's periodic inspection received damage at its power incoming unit due to the 2011 off the Pacific Coast of Tohoku Earthquake, which incurred loss-of-offsite-power. Immediately, two emergency diesel generators (D/G) automatically started, which supplied emergency system power. During eight days before the power incoming unit got provisional restoration, power supply through D/G continued. During this period, emergency measures for fuel and cooling water securement for D/G were taken, which forced unexperienced long-term load operation for D/G. This paper reports the change of plant conditions of Joyo, as well as the measures taken for keeping fuel and cooling water for continuing the operation of D/G. Since the main machines of the primary system and secondary system were functioning normally after the earthquake, the plant could be maintained in a stable state. As for the fuel for D/G, reduction in fuel consumption based on load suppression was implemented, and the cooperation of trading firms, related facilities, and local suppliers secured the fuel that supported the long-term operation of D/G. As for the cooling water, the cooperation of a self-defense fire brigade allowed to secure the required amount by utilizing the reserved water in the water tank for fire prevention. From these series of experiences of handling the plant, it was possible to extract the future challenges against the time when a massive earthquake or prolonged power failure has occurred. (A.O.)

  12. Numerical and experimental study of hydraulic dashpot used in the shut-off rod drive mechanism of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Narendra K., E-mail: nksingh_chikki@yahoo.com [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Badodkar, Deepak N. [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Singh, Manjit [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-07-01

    Highlights: • Hydraulic dashpot performance is studied numerically as well as experimentally. • Instantaneous pressure built-up in the dashpot is mainly contributing for damping of freely falling shut-off rod at the end of its travel. • At elevated temperature, dashpot pressure does not reduce in proportion to the reduction in viscosity. • ‘C’ grove in the dashpot shaft flattens the pressure peak and shifts it toward the end of operation. - Abstract: Hydraulic dashpot design for shut-off rod drive mechanism application in a nuclear reactor has been analyzed both numerically and experimentally in this paper. Finite element commercial code COMSOL Multiphysics 4.3 has been used for numerical analysis. Experimental validation has been done at two different cases. Experimental test set-ups and hydraulic dashpot constructions have been described in detail. Various combinations of dashpot oil viscosity and clearance thickness have been analyzed. Important experimental results are also presented and discussed. Pressure distributions in the dashpot chambers obtained from COMSOL are given for both the set-ups. Numerical and experimental results are compared. Dashpot designs have been qualified after detailed analysis and testing on full-scale test stations simulating actual reactor conditions (except radiation)

  13. An experimental and modeling study of propene oxidation. Part 1: Speciation measurements in jetstirred and flow reactors

    OpenAIRE

    Burke, Sinéad,; Metcalfe, Wayne; Herbinet, Olivier; Battin-Leclerc, Frédérique; Haas, Francis,; Santner, Jeffrey; Dryer, Frederick; J. Curran, Henry

    2014-01-01

    International audience Propene is a significant component of Liquefied Petroleum Gas (LPG) and an intermediate in the combustion of higher order hydrocarbons. To better understand the combustion characteristics of propene, this study and its companion paper present new experimental data from jet-stirred (JSR) and flow reactors (Part I) and ignition delay time and flame speed experiments (Part II). Species profiles from JSR experiments are presented and were obtained at near-atmospheric pre...

  14. Construction of an experimental simplified model for determining of flow parameters in chemical reactors, using nuclear techniques

    International Nuclear Information System (INIS)

    The development of a simplified experimental model for investigation of nuclear techniques to determine the solid phase parameters in gas-solid flows is presented. A method for the measurement of the solid phase residence time inside a chemical reactor of the type utilised in the cracking process of catalytic fluids is described. An appropriate radioactive labelling technique of the solid phase and the construction of an eletronic timing circuit were the principal stages in the definition of measurement technique. (Author)

  15. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy systems

    OpenAIRE

    Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This report contains two parts: (1) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (2) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showin...

  16. Experimental and calculation-theoretical justifications of design of hydrostatic bearings of MCP in reactor facility with HLMC

    International Nuclear Information System (INIS)

    Experimental and calculation and theoretical justification of normal operation of hydrostatic bearings (HSB) is conducted. The consideration is given to the procedures of calculation and recommendations on developing optimal designs of plain bearings in main circulation pumps working with high-temperature lead melt coolant for reactor circuit conditions. It is shown that chosen variants of HSB design after hydrothermal and tribotechnical investigations provide necessary characteristics of HSB on high-temperature lead melt

  17. Design and analysis on tritium system of multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

    Energy Technology Data Exchange (ETDEWEB)

    Ni Muyi, E-mail: nimuyi@mail.ustc.edu.cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China); Song Yong [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China); Jin Ming; Jiang Jieqiong [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Huang Qunying [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, Anhui, 230026 (China)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer A concept of the tritium system was designed for the FDS-MFX. Black-Right-Pointing-Pointer The system parameters were presented and discussed in detail. Black-Right-Pointing-Pointer A theoretical analysis of tritium recovery system has been made on the operation condition. Black-Right-Pointing-Pointer Three step TEP system was design and its permeating capacity was estimated. Black-Right-Pointing-Pointer The model of three-column ISS and the SDS was also carried out. - Abstract: As early application of fusion technology, the fusion-fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion-fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.

  18. Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    Highlights: ► A concept of the tritium system was designed for the FDS-MFX. ► The system parameters were presented and discussed in detail. ► A theoretical analysis of tritium recovery system has been made on the operation condition. ► Three step TEP system was design and its permeating capacity was estimated. ► The model of three-column ISS and the SDS was also carried out. - Abstract: As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.

  19. Development and experimental validation of a Monte Carlo simulation model for the Triga Mark II reactor

    OpenAIRE

    Chiesa,

    2014-01-01

    In recent years, many computer codes, based on Monte Carlo methods or deterministic calculations, have been developed to separately analyze different aspects regarding nuclear reactors. Nuclear reactors are very complex systems, which require an integrated analysis of all the variables which are intrinsically correlated: neutron fluxes, reaction rates, neutron moderation and absorption, thermal and power distributions, heat generation and transfer, criticality coefficients, fuel burnup, e...

  20. 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.

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

    International Nuclear Information System (INIS)

    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 heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  2. Development and optimization of neutron measurement methods by fission chamber on experimental reactors - management, treatment and reduction of uncertainties

    International Nuclear Information System (INIS)

    The main objectives of this research thesis are the management and reduction of uncertainties associated with measurements performed by means of a fission-chamber type sensor. The author first recalls the role of experimental reactors in nuclear research, presents the various sensors used in nuclear detection (photographic film, scintillation sensor, gas ionization sensor, semiconducting sensor, other types of radiation sensors), and more particularly addresses neutron detection (activation sensor, gas filling sensor). In a second part, the author gives an overview of the state of the art of neutron measurement by fission chamber in a mock-up reactor (signal formation, processing and post-processing, associated measurements and uncertainties, return on experience of measurements by fission chamber on Masurca and Minerve research reactors). In a third part, he reports the optimization of two intrinsic parameters of this sensor: the thickness of fissile material deposit, and the pressure and nature of the filler gas. The fourth part addresses the improvement of measurement electronics and of post-processing methods which are used for result analysis. The fifth part deals with the optimization of spectrum index measurements by means of a fission chamber. The impact of each parameter is quantified. Results explain some inconsistencies noticed in measurements performed on the Minerve reactor in 2004, and allow the improvement of biases with computed values

  3. 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

  4. International Thermonuclear Experimental Reactor U.S. Home Team Quality Assurance Plan

    Energy Technology Data Exchange (ETDEWEB)

    Sowder, W. K.

    1998-10-01

    The International Thermonuclear Experimental Reactor (ITER) project is unique in that the work is divided among an international Joint Central Team and four Home Teams, with the overall responsibility for the quality of activities performed during the project residing with the ITER Director. The ultimate responsibility for the adequacy of work performed on tasks assigned to the U.S. Home Team resides with the U.S. Home Team Leader and the U.S. Department of Energy Office of Fusion Energy (DOE-OFE). This document constitutes the quality assurance plan for the ITER U.S. Home Team. This plan describes the controls exercised by U.S. Home Team management and the Performing Institutions to ensure the quality of tasks performed and the data developed for the Engineering Design Activities assigned to the U.S. Home Team and, in particular, the Research and Development Large Projects (7). This plan addresses the DOE quality assurance requirements of 10 CFR 830.120, "Quality Assurance." The plan also describes U.S. Home Team quality commitments to the ITER Quality Assurance Program. The ITER Quality Assurance Program is based on the principles described in the International Atomic Energy Agency Standard No. 50-C-QA, "Quality Assurance for Safety in Nuclear Power Plants and Other Nuclear Facilities." Each commitment is supported with preferred implementation methodology that will be used in evaluating the task quality plans to be submitted by the Performing Institutions. The implementing provisions of the program are based on guidance provided in American National Standards Institute/American Society of Mechanical Engineers NQA-1 1994, "Quality Assurance." The individual Performing Institutions will implement the appropriate quality program provisions through their own established quality plans that have been reviewed and found to comply with U.S. Home Team quality assurance plan commitments to the ITER Quality Assurance Program. The extent of quality program provisions

  5. An experimental investigation of fission product release in SLOWPOKE-2 reactors

    International Nuclear Information System (INIS)

    Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

  6. Power transient analyses of experimental in-reflector devices during safety shutdown in Jules Horowitz Reactor (JHR)

    Energy Technology Data Exchange (ETDEWEB)

    Camprini, P. C.; Sumini, M. [Univ. of Bologna (Italy); Artioli, C. [National Agency for New Technologies, Energy and Sustainable Economic Development ENEA (Italy); Gonnier, C.; Pouchin, B.; Bourdon, S. [Atomic Energy Commission CEA (France)

    2012-07-01

    The Jules Horowitz Reactor (JHR) is designed to be a 100 MW material testing reactor (MTR) and it is expected to become the reference facility in the framework of European nuclear research activity. As the core neutron spectrum is quite fast, several experimental devices concerning fuel studies have been conceived to be placed in the reflector in order to exploit a proper thermal neutron flux irradiation. Since the core power is relatively high, the neutronic coupling between the reactor core and the reflector devices has to be taken into account for different rod insertions. In fact the thermal power produced within the fuel samples is considerable. Heat removal during shutdown is a main topic in nuclear safety and it is worth to analyse thermal power transients in fuel samples as well. Here a thermal hydraulic model for JHR core is proposed aiming at a simple and representative description as far as reactivity feedbacks are concerned. Then it is coupled with a neutronic pointwise kinetics analysis by means of the DULCINEE code to compute core power transient calculations. Moreover, some reflector-core coupling evaluations are performed through Monte Carlo method using the TRIPOLI 4.7 code. The JHR equilibrium cycle is considered with respect to four fuel compositions namely Beginning of Cycle (BOC), Xenon Saturation Point (XSP), Middle of Cycle (MOC) and End of Cycle (EOC). Then thermal power transients in the experimental reflector devices are evaluated during safety shutdowns and they are verified for all these cycle steps. (authors)

  7. An experimental substantiation of the design functions imposed on the additional system for passively flooding the core of a VVER reactor

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from a research work on experimentally substantiating the serviceability of the additional system for passively flooding the core of a VVER reactor from the second-stage hydro accumulators are presented.

  8. Integrated detoxification methodology of hazardous phenolic wastewaters in environmentally based trickle-bed reactors: Experimental investigation and CFD simulation

    Energy Technology Data Exchange (ETDEWEB)

    Lopes, Rodrigo J.G., E-mail: rodrigo@eq.uc.pt [Centro de Investigacao em Engenharia dos Processos Quimicos e Produtos da Floresta (CIEPQPF), GERSE - Group on Environmental, Reaction and Separation Engineering, Department of Chemical Engineering, University of Coimbra, Rua Silvio Lima, Polo II - Pinhal de Marrocos, 3030-790 Coimbra (Portugal); Almeida, Teresa S.A.; Quinta-Ferreira, Rosa M. [Centro de Investigacao em Engenharia dos Processos Quimicos e Produtos da Floresta (CIEPQPF), GERSE - Group on Environmental, Reaction and Separation Engineering, Department of Chemical Engineering, University of Coimbra, Rua Silvio Lima, Polo II - Pinhal de Marrocos, 3030-790 Coimbra (Portugal)

    2011-05-15

    Centralized environmental regulations require the use of efficient detoxification technologies for the secure disposal of hazardous wastewaters. Guided by federal directives, existing plants need reengineering activities and careful analysis to improve their overall effectiveness and to become environmentally friendly. Here, we illustrate the application of an integrated methodology which encompasses the experimental investigation of catalytic wet air oxidation and CFD simulation of trickle-bed reactors. As long as trickle-bed reactors are determined by the flow environment coupled with chemical kinetics, first, on the optimization of prominent numerical solution parameters, the CFD model was validated with experimental data taken from a trickle bed pilot plant specifically designed for the catalytic wet oxidation of phenolic wastewaters. Second, several experimental and computational runs were carried out under unsteady-state operation to evaluate the dynamic performance addressing the TOC concentration and temperature profiles. CFD computations of total organic carbon conversion were found to agree better with experimental data at lower temperatures. Finally, the comparison of test data with simulation results demonstrated that this integrated framework was able to describe the mineralization of organic matter in trickle beds and the validated consequence model can be exploited to promote cleaner remediation technologies of contaminated waters.

  9. Integrated detoxification methodology of hazardous phenolic wastewaters in environmentally based trickle-bed reactors: Experimental investigation and CFD simulation.

    Science.gov (United States)

    Lopes, Rodrigo J G; Almeida, Teresa S A; Quinta-Ferreira, Rosa M

    2011-05-15

    Centralized environmental regulations require the use of efficient detoxification technologies for the secure disposal of hazardous wastewaters. Guided by federal directives, existing plants need reengineering activities and careful analysis to improve their overall effectiveness and to become environmentally friendly. Here, we illustrate the application of an integrated methodology which encompasses the experimental investigation of catalytic wet air oxidation and CFD simulation of trickle-bed reactors. As long as trickle-bed reactors are determined by the flow environment coupled with chemical kinetics, first, on the optimization of prominent numerical solution parameters, the CFD model was validated with experimental data taken from a trickle bed pilot plant specifically designed for the catalytic wet oxidation of phenolic wastewaters. Second, several experimental and computational runs were carried out under unsteady-state operation to evaluate the dynamic performance addressing the TOC concentration and temperature profiles. CFD computations of total organic carbon conversion were found to agree better with experimental data at lower temperatures. Finally, the comparison of test data with simulation results demonstrated that this integrated framework was able to describe the mineralization of organic matter in trickle beds and the validated consequence model can be exploited to promote cleaner remediation technologies of contaminated waters. PMID:21377790

  10. Experimental determination of spectral indices by scanning of fuel rod in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    In this work, the spectral indexes 28ρ* and 25δ*, and gamma efficiency factor in the IPEN/MB-01 reactor were determined experimentally employing a rod scanning technique. In the case of 28ρ*, 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 235U remain in the determination of the 25δ*. 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)

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

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    International Nuclear Information System (INIS)

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs

  14. Experimental investigation into the effects of coolant additives on boiling phenomena in pressurized water reactors

    International Nuclear Information System (INIS)

    This study investigates the effects of coolant additives like boric acid on boiling phenomena in pressurized water reactors under conditions as realistic as possible. The effects covered range from subcooled boiling to critical boiling conditions (CHF). The focus of this project lies on flow boiling with up to 40 bar and 250 °C in order to generate a data basis for a possible extrapolation to reactor conditions. The results of the experiments are used to implement and validate new models into CFD-Codes in context to a nationwide German joint research project with the specific aim of improving CFD boiling-models. (author)

  15. 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.

  16. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  17. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Henrique F.A.; Ferreira, Andrea V., E-mail: hfam@cdtn.br, E-mail: avf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  18. 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…

  19. 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, numerica

  20. 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...

  1. Experimental Study on the Hydrodynamics and Mass Transfer Characteristics of Airlift Loop Reactors(ALR)

    Institute of Scientific and Technical Information of China (English)

    Tang Lixin; Han Pingfang; Lu Xiaoping

    2007-01-01

    The promoting effect of ultrasonic wave on the hydrodynamics and mass transfer characteristics of the airlift loop reactor was studied. The effect of the airlift reactor and ultrasonic wave on the reactor's gas holdup, liquid circulation velocity, mixing time and overall volumetric mass transfer coefficient respectively with and without the presence of ultrasonic wave is empathetically examined and compared. The experiment has proven that the incorporation of ultrasonic wave has no effect on the gas holdup but has the tendency to gradually decrease the liquid circulation velocity and increase the overall volumetric mass transfer coefficient; the effect on the mixing time is relatively complex. At low gas velocity, low powered ultrasonic wave promotes the radial mixing of fluid; with the increase of ultrasonic power, ultrasonic vibration obstructs the radial mixing of fluid. Therefore, there exists an optimal ultrasonic power. Moreover, the effect of ultrasonic wave on the mixing time gradually decreases with the increase of the superficial gas velocity. Correlations were also proposed for the hydrodynamics and mass transfer characteristics of the reactor.

  2. ITER: design, construction and operation of te first experimental nuclear fusion reactor

    International Nuclear Information System (INIS)

    The ITER project's mission is to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes and its now under construction at Saint Paul-Lez-Durance (France). The ITER reactor is based on the 'tokamak' concept of plasma magnetic confinement, in which the fusion (deuterium-tritium) fuel is contained in a doughnut-shaped vessel. The plasma is kept away from the vessel walls by strong magnetic fields produced by superconducting coils surrounding the vessel and by an electrical current driven in the plasma. The ITER reactor is designed to generate 500 MW of fusion power for periods of 300 to 500 seconds with a fusion power multiplication factor, Q, of at least 10 (Q ≥ 10). ITER will also aim at demonstrating long fusion power production pulses, of at least 1000 seconds, with a fusion power multiplication factor of 5 and, ultimately, of 1 hour duration (only limited by hardware design limits) when full non-inductive operation is demonstrated. The presentation will cover the following aspects of the ITER reactor design, construction and foreseen operation: a) The basis for power production by magnetic confinement tokamak fusion reactors. b) The main features of the ITER tokamak reactor design. c) The key design principles of the ITER tokamak and of the key ancillary systems required for the operational scenarios considered to achieve the project's mission. d) The design and progress in qualification and manufacturing of the key ITER tokamak infrastructure, tokamak components and ancillary systems. e) The operational plan from the initial commissioning phase, through operation with non-nuclear hydrogen-helium plasmas to nuclear operation with deuterium-tritium plasmas and the demonstration of high Q fusion power. (author)

  3. IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

    International Nuclear Information System (INIS)

    Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

  4. A review of theoretical and experimental studies underlying the thermal-hydraulic design of fast reactor fuel elements

    International Nuclear Information System (INIS)

    The economic performance of fast reactors is closely linked to the achievable burn-up of heavy atoms, that is to the endurance life of the fuel pins. The safety case must also be concerned with the integrity of the cladding, since this is the primary containment envelope for fission products. It is thus important to ensure that cladding temperatures during reactor operation are limited to levels which incur no serious impairment of mechanical properties. The function of thermal-hydraulic analysis is to provide fuel element designers with the means of achieving this objective. This paper reviews the theoretical approaches which have been developed and applied in the UK in the design of LMFBR fuel and breeder sub-assemblies, control rods and experimental clusters. It also presents results of experimental studies undertaken to develop a better understanding of coolant flow distribution and mixing problems in these components, and to provide essential data for computer codes. Problem areas in this field are highlighted, particularly the difficulties arising due to irradiation induced distortions. Reference is made to the experimental and theoretical developments which are in progress, or may be required, to provide adequate predictions of fuel pin temperatures at high burn-up. (author)

  5. FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

    International Nuclear Information System (INIS)

    The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

  6. Executive mechanisms control in structure of experimental plants TEXT, DBR of the IBR-2, IBR-30 reactors

    International Nuclear Information System (INIS)

    Task of executive mechanisms control in the structure of experimental plant TEXT, DBR of the IBR-2, IBR-30 reactors is made up. The application zone of the CAMAC control block (BIUM) on the basis of the K1801VM1 microprocessor is extended: for different types of executive mechanisms with control pulse; for different types of executive mechanisms with the movement between control points. The block of control and monitoring information of executive mechanisms is represented by the text file of PK. 6 refs.; 5 figs.; 1 tab

  7. Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

    International Nuclear Information System (INIS)

    A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

  8. Experimental studies on AC dielectric barrier corona discharge reactors for NO{sub x}/SO{sub x} decompositions

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y.H.; Hong, S.H. [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    1997-12-31

    Recently the interest in the AC dielectric barrier corona discharge (DBCD) is rapidly growing because of their successful application to the decomposition of air pollutants such as NO{sub x}/SO{sub x}. In this work, two types of DBCD reactors, cylinder-wire and pins-plate, are designed and experimented, and their decomposition efficiencies of NO{sub x} and SO{sub x} are compared. The powered electrodes of either tungsten wire or pins having a radius of curvature of 1 mm are fixed at 20 mm away from dielectric barriers of pyrex tube or acryl plate with a thickness of 2 mm. Stainless steel tube and plate of 100 mm in length are used as a grounded electrode of both type reactors, respectively. The gas treatment time is about 1 sec at a flow rate of 8 1/min in a discharge volume of about 120 cm{sup 3}. An AC high-voltage transformer, simulating pollutant gases containing 500-ppm NO or SO{sub 2}, and a gas analyzer are employed in the decomposition experiments for finding out an optimal reactor type and operating parameters.Decomposition efficiencies are measured by varying the amplitudes and frequencies of applied voltage and the flow rates and compositions of simulating pollutant gases. The experimental results show that the cylinder-wire type is more efficient in decomposing NO and So{sub 2} than pins-plate one.

  9. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  10. 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.)

  11. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    International Nuclear Information System (INIS)

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components

  12. Calibration of the enigma code for Finnish reactor fuel with support from experimental irradiations

    International Nuclear Information System (INIS)

    Assessment by VTT of the ENIGMA fuel performance code, the original version by Nuclear Electric plc of the UK amended by a set of WWER specific materials correlations, is described. The given examples of results include analyses for BWR 9 x 9 fuel, BWR fuel irradiated in the reinstrumented test of an international Riso project, pre-characterized commercial WWER fuel irradiated in Loviisa reactor in Finland, and instrumented WWER test fuel irradiations in the MR reactor in Russia. The effects of power uncertainty and some model parameters are discussed. Considering the fact that the described cases all mean prototypic application of the code, the results are well encouraging. The importance of the accuracy in temperature calculations is emphasized. (author). 2 refs, 12 figs, 1 tab

  13. Experimental Study of a Stoppage Natural Circulation during a Nuclear Heating Reactor LOCA

    Institute of Scientific and Technical Information of China (English)

    博金海; 张佑杰; 姜胜耀

    2001-01-01

    The 5MW nuclear heating reactor is an integral naturalcirculation reactor. The rupture of the coolant penetrating tube is a typical accident causing coolant loss. When the water level drops down to the upper edge of the inlet of the heat exchanger, the natural circulation stops. This influences the core cooling and the stability of the main loop. A series of tests showed that there is a stable drop of pressure, and the heated element temperature is not too high to cause burnout. But the backward flow or flow oscillation in the primary coolant circuit occurs when the flow breaks completely in the end. The whole flow process is described and the mechanism is discussed.

  14. Study on improvement of core management and irradiations field characterization methods of the experimental fast reactor Joyo (Thesis)

    International Nuclear Information System (INIS)

    This thesis describes the research study to develop the core management method and irradiation field characterization method of the experimental fast reactor Joyo. Improvements of the methods through comparison with measured data from the reactor core physics performance tests of Joyo and post irradiation examination (PIE) of tests conducted in the Joyo irradiation test facility complex are also described. There are eight chapters. Chapter 1 describes the objectives of this study, along with a brief history of the Joyo test reactor and an explanation of the role and importance of developing the sodium cooled fast breeder reactor (FBR) in Japan from the view point of providing the future energy source. Chapter 2 explains the core management method of the Joyo Mark-II irradiation core, which had been modified from the first Mark-I breeder core. The core management method modifications of Joyo included changing the refueling scheme by employing an in-out fuel shuffling method and re-examination of the thermal design margin of the driver fuel by reducing the hot spot factor based on the evaluation of the Joyo Mark-II core and plant performance tests. Chapter 3 describes the development of improved methods for evaluating the neutron and gamma flux distributions by including energy spectrum information in order to meet the requirements for their accuracy. These developments included modifying the analytical method and developing the new neutron dosimetry method of helium accumulation fluence monitor (HAFM). These improvements were validated by comparison with the measured reaction rates obtained by the conventional multiple foil activation method. Chapters 4 and 5 describe the design of the upgrade of the Joyo core and cooling system, called the Mark-III project, in order to increase the neutron flux 1.3 times higher than the original design maximum of the Joyo Mark-II core. The modified Mark-III core and plant performance test evaluations that were used to validate the

  15. Experimental and modelling evaluation of an ammonia-fuelled microchannel reactor for hydrogen generation / Steven Chiuta

    OpenAIRE

    Chiuta, Steven

    2015-01-01

    In this thesis, ammonia (NH3) decomposition was assessed as a fuel processing technology for producing on-demand hydrogen (H2) for portable and distributed fuel cell applications. This study was motivated by the present lack of infrastructure to generate H2 for proton exchange membrane (PEM) fuel cells. An overview of past and recent worldwide research activities in the development of reactor technologies for portable and distributed hydrogen generation via NH3 decomposition wa...

  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 studies on in-bundle ECCS injection for Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    The Advanced Heavy Water Reactor (AHWR) being designed at BARC is an innovative reactor with Thorium utilization as its major objective. It has many advanced passive safety features. One such feature is passive injection of emergency coolant after postulated Loss of Coolant Accident (LOCA). A novel feature of this injection scheme is that the injection does not take place in the header/plenum as in other reactors, but directly in to the bundle. For this purpose, the fuel cluster incorporates a central water rod which communicates with the ECCS header. The water rod extends along full length of the fuel cluster. In event of LOCA in the Main Heat Transport (MHT) system, ECC water flows from the accumulator to the water rod through ECCS header. The water flows into the bundle through holes in the water rod. The AHWR fuel cluster has fuel pins arranged in three concentric rings (of 12, 18 and 24 pins) around the central rod. While it is ensured that water does reach the fuel cluster, whether it reaches the outer ring of pins is needs investigation as the pins are closely spaced (1-3 mm gap between adjacent rods). The objective of the present experiments is to determine under what conditions (ECC flow and decay heat), the ECC water is able to rewet and cool all the fuel pins. The experiments have been done in a short, instrumented fuel bundle simulating the geometry of the AHWR fuel cluster

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

    International Nuclear Information System (INIS)

    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-α 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 and calculational investigation of longitudinal-and-transverse coolant flow in the fuel assembly of a gas cooled reactor

    International Nuclear Information System (INIS)

    A method for calculating fields of rates temperature and pressure of gas coolant in two-dimensional setup in a gas-cooled reactor fuel assembly including an inlt monifold porous energy-release medium and outlet monifold is considered. From the view of structure the fuel assembly coaxially located porous cylinders with spherical fuel elements between them. An experimental large-scale facility permitting to perform a set of necessary measurements is described. Fields of rtes in the inlet manifold were measured as well as pressure distributions on external (imprenetrable) and internal (perforated) walls of the inlet manifold on its end face wall, on the perforated wall of the outlet manifoled. Experimental and calculational results have been compared

  20. An experimental parametric study of the high pressure melt ejection from two different scale reactor cavity models

    International Nuclear Information System (INIS)

    A parametric study of the high pressure melt ejection(HPME) from two small-scale(1/25th and 1/41st) transparent reactor cavity models of the YoungGwang(unit 1 and 2) has been performed. Wood's metal and water have been used as melt simulants while high pressure nitrogen and carbon dioxide are used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. Experimental data for the fraction of melt simulant retained in the cavity model(Yf) during a postulated scenario of the HPME from PWR pressure vessel have been obtained as a function of various test parameters. These data have been used to develop a correlation for Yf that fits all the data(a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least-squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Yf are also determined based on the correlation obtained here and experimental evidence. (Author)

  1. An experimental low-pressure facility to study boron transients in the pressurizer of an integral modular nuclear reactor

    International Nuclear Information System (INIS)

    Small and medium size modular reactors offer many advantages when compared with typical nuclear plants in various circumstances, such as offering greater simplicity of design, economy of mass production, and reducing siting costs. The integral configuration is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. However, for this configuration there is no spray system for boron homogenization, which may cause power transients. Thus, it is necessary to investigate boron mixing. The Federal University of Pernambuco (UFPE), in a partnership with the Regional Center of Nuclear Sciences of Northeast (CRCN-NE) and the Nuclear Engineering Institute (IEN/CNEN-RJ), is developing a project that aims to analyze transients in a compact modular integral reactor. This analysis will be made by using the data obtained from one experimental bench that is mounted at CRCN-NE. A study accomplished in 2012 using a simplified bench (built in reduced scale with a test section manufactured with transparent acrylic) showed that it was possible to obtain preliminary experimental results for the boron homogenizing process. (author)

  2. Experimental Evidence of a Double Layer in a Large Volume Helicon Reactor

    OpenAIRE

    Sutherland, O; Charles, C.; Plihon, N; Boswell, Rod

    2005-01-01

    International audience The self-consistently generated current-free electric double layer (DL) is shown to scale up with the source tube diameter and appears not to be affected by rf driving frequency and changes in reactor geometry. This Letter presents the first simultaneous measurements of local plasma potential and beam energy as a function of axial position. The DL is shown to be no more than 5 mm thick (20 D lengths) and positioned just downstream of the maximum in the magnetic field...

  3. Model to simulate the fission-product transport process in the Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    So, B.Y.C.

    1979-01-01

    When fission products are released from a cladding breach in EBR-II, they mix turbulently with the sodium in the core, in the upper plenum and in the intermediate heat exchanger. Eventually the fission products are discharged 12 to 13 s later into the primary tank. Fission gases migrate upward through a 9-ft layer of sodium and enter the cover gas. Loss of fission gas is due to decay, leakage of cover gas, cold trapping of iodine and bromine parents. Depending on the reactor operation requirement, it may purge with fresh argon. The assumptions made and differential equations used to develop a model for such transport are presented.

  4. Problems in experimental and mathematical investigations of the accidental thermalhydraulic processes in RBMK nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nigmatulin, B.I.; Tikhonenko, L.K. [Engineering Centre (EREC) for Nuclear Plants Safety, Electrogorsk (Russian Federation); Blinkov, V.N. [Aviation Institute, Kharkov (Ukraine)] [and others

    1995-09-01

    In this paper the thermalhydraulic scheme and peculiarities of the boiling water graphite-moderated channel-type reactor RBMK are presented and discussed shortly. The essential for RBMK transient regimes, accidental situations and accompanying thermalhydraulic phenomena and processes are formulated. These data are presented in the form of cross reference matrix (version 1) for system computer codes verification. The paper includes qualitative analysis of the computer codes and integral facilities which have been used or can be used for RBMK transients and accidents investigations. The stability margins for RBMK-1000 and RBMK-1500 are shown.

  5. Research reactors

    International Nuclear Information System (INIS)

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

  6. An Experimental Study on the Breakup of Simulant Melt Jet Released from the Submerged Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    The potential risk of explosive molten fuel coolant interactions (FCI, steam explosion) has drawn substantial attention in the safety analysis of reactor severe accidents. The steam explosion intensity is largely dependent upon the degree of volumetric fractions of melt droplets and steam in the fuel-coolant mixture. The rate of melt jet breakup and droplet sizes are, therefore, the key physical parameters in the analysis of FCIs. An ex-vessel steam explosion may occur when the core melt is released from the failed reactor vessel lower head into the water-filled reactor cavity. The water level in the cavity can be either below or above the reactor vessel lower head depending on the severe accident management strategy. The former, a partially-filled cavity with free-fall space for the melt jet, has been the major condition for the steam explosion studies in the past. An In-Vessel Retention by External Vessel Cooling (IVR-ERVC), however, requires the water level in the cavity be above the reactor vessel lower head so that the vessel can be completely submerged in coolant water. In this case, the melt jet falls in liquid water without free fall. The jet breakup behavior in such a condition has been rarely studied, nor well identified. In this work, jet breakup of the melt released from the submerged vessel has been experimentally investigated using stimulant melt of Woods metal. The initial melt temperature was set below the boiling point of water so that only the hydrodynamic mechanism of jet breakup can be identified. High-speed videos were taken to visualize the jet breakup behavior and the post-test debris were collected and sieved to obtain debris size distributions. Non-boiling liquid jet breakup experiment was conducted using 50 mm-diameter Woods metal jet released from a vessel submerged in water pool. The post-test debris was sieved and the debris size distributions were obtained. The size of the largest mass of the debris shows a fair agreement with the

  7. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  8. 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.

  9. 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.

  10. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  11. 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

  12. Hydraulic Experiment for Simulative Assemblies of Blanket Assembly and Np Transmutation Assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    CHENG; Dao-xi; QI; Xiao-guang; ZHAI; Wei-ming; YANG; Bing; ZHOU; Ping

    2013-01-01

    The out-of reactor hydraulic experiment of fast reactor assembly is one of the important experiments in the process of the development of the fast reactor assembly.In this experiment,the size of the throttling element in the foot of the assembly is decided which is fit for the flow division in the reactor and the

  13. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  14. Review of the n_TOF experimental program for Reactor Applications

    Directory of Open Access Journals (Sweden)

    Guerrero C.

    2013-03-01

    Full Text Available The n_TOF facility at CERN is devoted mainly to the measurement of neutron-induced reaction cross section of interest for Nuclear Technologies, Astrophysics and Basic Physics. In particular, the list of measurements carried out during the 2nd Phase of experiments n_TOF-Ph2 (2009-2012 includes a significant number of capture and fission experiments on actinides which are considered key for the further development of nuclear reactors. This contribution will contain a description of all these experiments, some of which will be discussed in detail. The future of the n_TOF facility will be also addressed; in particular, the new vertical neutron beam line with a flight path of only 20 m will be presented and the expected performance discussed in detail.

  15. Experimental investigations on load reduction in the pressure suppression system of boiling water reactors

    International Nuclear Information System (INIS)

    For the load specification of pressure suppression systems in boiling water reactors the periodic pressure pulses from a condensation phenomenon, called chugging, are of great importance. The research indicates, that the chugging mechanism is mainly induced by the BORDA-effect at the sharp edge of the vent pipe outlet. Based on these insights, simple vent pipe outlet mitigators are developed and tested, which effect in a passive mode a significant reduction of the dynamic pressure pulses from this condensation phenomenon. The results also yield the proof of multivent effect, of time window for single chugging event occurrence at a multivent configuration and the assurance of the reproducibility of this dynamic condensation phase. (orig.)

  16. Safety analysis of superconducting poloidal magnets for a JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    The design of superconducting poloidal magnets was restudied concerning safety. The magnetmotive force of the superconducting air core transformer was reduced from 60 MAT to 48 MAT by shortening the operation period of reactor, so that the maximum repeated tensile stress of structural material (304SS) caused by electromagnetic force was lowered to the level of 24 kg/mm2 and stability margin of the superconductor was increased. The deterioration of thermal insulation in the vacuum jacket was studied, and safety valves were designed to cope with it. For repair, dissasembling the poloidal magnet was also studied in relation of the toroidal magnet. Analysis of general failure of the superconducting poloidal magnet was made by event tree and fault tree techniques. (author)

  17. Experimental and analytical studies of passive shutdown heat removal from advanced LMRs [liquid metal reactors

    International Nuclear Information System (INIS)

    A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary Cooling System (RVACS). The NSTF operates in either a uniform heat flux mode and a uniform temperature mode at the air/guard vessel interface. Analysis of the RVACS performance data indicates excellent agreement with pretest analytical predictions. Correlation analysis presents the heat transfer data in a form suitable for use in LMR design and verification of analytical studies

  18. Modelling of Multi-Physics Phenomena in Fast Reactor Design: Safety and Experimental Validation

    International Nuclear Information System (INIS)

    The paper provides a cursory look at current approaches in numerical modelling and simulation of typical multi-physics phenomena of concern relevant to the sodium cooled fast reactor design and safety. Emphasis is placed on the methods that are in practice and their verification and validation programmes, including for those of fluid–structure thermal interactions due to thermal striping, thermodynamics of sodium–water chemical reactions, multi-component and multi-phase flows in the fuel degradation and core meltdown phases. Several of the numerical simulations of these phenomena are shown with verification and validation programs that employ not only separate effect small scale experiments of clean geometry but also for large scale integral tests or mock-up experiments. The last part of this paper will be spent on discussions on a more quantitative validation basis with identification of errors and/or uncertainties based on the Bayesian rule. (author)

  19. Activation analysis on reactor and 14 MeV neutrons in experimental study of oxoplatinum

    International Nuclear Information System (INIS)

    A procedure of the instrumental neutron activation analysis of biomedical material on reactor and 14 MeV neutrons was described for determination of the platinum, silicon, chlorine, phosphorus, sodium and zinc content. The procedure was tested on rat small intestinal samples within the interval of 10 min - 30 days after a single i.v. administration of oxoplatinum (15 mg per 1 kg of the animal body mass). Platinum redistribution with relative concentration maximum on the 15th day was noted. Changes in the phosphorus, chlorine, sodium and silicon content within the interval of 3-30 days were suggestive of a possible toxic effect of oxoplatinum metabolites during this period. In a clinical application of oxoplatinum one should take into account a possible toxic effect of metabolites in a long-term period after repeated administration of the agent

  20. Safety properties of China experimental fast reactor%中国实验快堆的安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤

    2011-01-01

    Sodium cooled fast reactor possesses some inherent safety properties, thanks to sodium perfect thermo-physical characteristics. In the same time sodium leakage inducing sodium fire or sodium-water reaction of industrial incidents, from sodium containing systems could not be excluded due to it is alkali metal. It is presented in the paper, that the safety of the China experimental fast reactor(CEFR)has meet the safety demands of Generation ]V due to the inherent safety characteristics have been realized, some passive safety systems, like passive decay heat removal system based on natural convection and circulation and active safety measures have been equipped. As for the large sized fast reactor with high breeding feature which induces positive sodium bubble effect, it is needed to develop passive shut-down systems to keep the safety targets of Generation IV.%钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保证高的增殖而会有正的钠空泡效应,需要开发非能动停堆系统以保持第Ⅳ代安全目标.

  1. 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.

  2. 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.

  3. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shripad T. Revankar; Seungmin Oh

    2003-09-30

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  4. Analytical and Experimental Study of The Effects of Non-Condensable in a Passive Condenser System for The Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    The main goal of the project is to study analytically and experimentally condensation heat transfer for the passive condenser system relevant to the safety of next generation nuclear reactor such as Simplified Boiling Water Reactor (BWR). The objectives of this three-year research project are to: (1) obtain experimental data on the phenomenon of condensation of steam in a vertical tube in the presence of non-condensable for flow conditions of PCCS, (2) develop a analytic model for the condensation phenomena in the presence of non-condensable gas for the vertical tube, and (3) assess the RELAP5 computer code against the experimental data. The project involves experiment, theoretical modeling and a thermal-hydraulic code assessment. It involves graduate and undergraduate students' participation providing them with exposure and training in advanced reactor concepts and safety systems

  5. Theoretical and experimental research of natural convection in the core of the high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    The physical model of the developed THERMIX-2D-code for computing thermohydraulic behaviour of the core of high temperature pebble bed reactors is verified by experiments with natural convection flow. Such fluid flow behaviour can be of very high importance for the real reactor in the case of natural heat removal decay. The experiments are performed in a special set up testing-stand with pressures up to 30 bars and temperatures up to 3000C by using air and helium as fluid. In comparison with the experimental data the numerical results show that a good and useful simulation is given by the program. Pure natural convection flow in packed pebble beds is calculated with a very high degree of reliability. The investigation of flow stability demonstrate that radial-symmetric relations are not given temporarily when national convection is overlayed by forced convection flow. In the discussion it is explained when and to what extent the program leds to useful results in such situations. The test of the effective heat conductivity lambdasub(eff) results in an improvement of the lambdasub(eff)-data used so far for temperatures below 13000C. (orig.)

  6. ICARUS.2: a kinetics analysis code for the experimental very high temperature gas-cooled reactor plant

    International Nuclear Information System (INIS)

    The report describes analytical models, computational methods and physical properties, which are employed in the computational code ICARUS.2. The code simulates the transient behavior of the experimental Very High Temperature Reactor plant. Transient responses of the plant to various disturbances are analyzed as well as of the typical components in the plant. The code has the enough number of modeled components to cover the two-loop system of the real plant. Each loop consists of three internal coolant loops which are the primary and the secondary helium coolant loops and the ternary water-steam coolant loop. Further, the code can model any structure of plant coolant system which might cause the significant effect on the transient behavior. (author)

  7. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling

  8. ITER: The International Thermonuclear Experimental Reactor and the nuclear weapons proliferation implications of thermonuclear-fusion energy

    CERN Document Server

    Gsponer, A; Gsponer, Andre; Hurni, Jean-Pierre

    2004-01-01

    This paper contains two parts: (I) A list of "points" highlighting the strategic-political and military-technical reasons and implications of the very probable siting of ITER (the International Thermonuclear Experimental Reactor) in Japan, which should be confirmed sometimes in early 2004. (II) A technical analysis of the nuclear weapons proliferation implications of inertial- and magnetic-confinement fusion systems substantiating the technical points highlighted in the first part, and showing that while full access to the physics of thermonuclear weapons is the main implication of ICF, full access to large-scale tritium technology is the main proliferation impact of MCF. The conclusion of the paper is that siting ITER in a country such as Japan, which already has a large separated-plutonium stockpile, and an ambitious laser-driven ICF program (comparable in size and quality to those of the United States or France) will considerably increase its latent (or virtual) nuclear weapons proliferation status, and fo...

  9. An experimental study of the partial oxidation of ethane to ethylene in a shallow fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    DANICA BRZIC

    2007-02-01

    Full Text Available The partial catalytic oxidation of ethane to ethylene was investigated experimentally in a shallow fluidized bed. The performaces of two catalyst types, pure g‑Al2O3 and V2O5/ g-Al2O3 particles 1.8 mm in diameter, were analyzed. A pilot fluidized bed reactor with rectangular cross-section of 100mm´100mm was used. The experiments were carried out under atmospheric pressure in a dilute system under oxygen excess conditions. V2O5/g-Al2O3 showed good catalytic performances regarding ethylene selectivity. The influence of the temperature (in the range of 400–600 °C and the contact time (in the range of 35 – 85 kg sm-3 on the conversion of ethane and the selectivity to ethylene was analyzed. The highest yield of ethylene was 18 %.

  10. Experimental possibilities and fast neutron dose map of the fast neutron fields at the RB reactor facility

    International Nuclear Information System (INIS)

    The RB is an unshielded, zero power nuclear facility with natural and enriched uranium fuel (2% and 80%) and D2O as moderator. It is possible to create different configurations of non-reflected and partially reflected critical systems and to make experiments in the fields of thermal neutrons. The fields of fast neutrons with 'softened' fission spectrum are made by modifying the system: modified experimental fuel channel EFC, coupled fast-thermal system in two configurations CFTS-1 and CFTS-2, coupled fast-thermal core HERBE. The intermediate and fast neutron absorbed doses in fast neutron fields are given. In first configuration of RB reactor it was almost impossible to perform dosimetric and other experiments. By creating these fields, with in our circumstances available fuel elements, the possibilities for different experiments are greatly improved. Now we can irradiate food samples, soil samples, electronic devices, study material properties, perform various dosimetry experiments, etc. (1 tab.)

  11. Experimental and computational studies on the natural circulation characteristics of a small- and medium-sized reactor, REX-10

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Byeong Il

    2012-02-15

    The rapid economic development in many countries has led to a growing need for energy. To meet the demand, the nuclear power generation is emerging as a viable alternative because it emits much smaller amount of carbon dioxide as compared to conventional fossil fuel plant produces high power density energy. Because of the latter attractiveness, many large-sized nuclear reactors ({approx}1000 MWe/unit) for electricity generation have been constructed around the world. Recently, small- and medium-sized nuclear reactors (SMRs) receive attention because of its effective adaptability to diversified energy demands. Several attractiveness of the SMR is worth to report here. First, the SMR is a viable alternative for the developing countries that have inferior electricity grids. Second, the SMR may effectively prevent the Large-Break Loss of Coolant Accident (LBLOCA) as it adopts the integral-type reactor. Finally, the SMR is used for various applications - desalination, district heating, small-scale power generation, ship propulsion, to mention a few. These attractiveness provides reasonable justification to construct the SMRs such as SMART, IRIS, CAREM, and KLT-40S. As such, the Regional Energy Research Institute for Next Generation (RERI) has been developing a SMR for regional energy supply, REX-10 since 2005. The REX-10 is the integral-type reactor and relies on natural circulation to improve passive safety. In this study, the natural circulation behavior and capability of the REX-10 is investigated experimentally as well as numerically. To evaluate the thermal-hydraulics of the REX-10, two experimental facilities . RTF and NACTER . were designed and fabricated using the scaling laws proposed by Ishii et al. Both experimental facilities consist of main components of the core, heat exchanger, hot legs, pressurizer, and chillers; however, both experimental facilities have different heat exchanger systems because two types of the heat exchanger are being considered. To

  12. Experimental subcritical reactivity determinations employing APSD measurements with pulse mode detectors in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Lee, Seung Min; Diniz, Ricardo; Jerez, Rogerio, E-mail: asantos@ipen.br, E-mail: smlee@ipen.br, E-mail: rdiniz@ipen.br, E-mail: rjerez@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This work aims to determine experimentally the subcritical reactivity levels of several configurations of the IPEN/MB-01 reactor in an approach based on the subcritical kinetic model developed by Gandini and Salvatores. The procedure employs the measurements of the APSD (Auto Power Spectral Density) using pulse mode detectors. The proposed approach is based only on measured quantities such as counting rates and the parameters arising from the least square approach of the APSD. Other difficult quantity such as detector efficiencies is not needed in the method. Several measurements of APSD were performed in varying degrees of sub-criticality (up to around -7000 pcm). The APSD data were least-square fitted to get the prompt decay mode (α). Beside the startup source, an external neutron sources of Am-Be was installed near the core in order to improve neutron count statistics. The final experimental results are of very good quality. The experiment shows clearly that the classical one point kinetic theory cannot describe the measured reactivity. MCNP K{sub eff} results were compared to the corresponding experimental results. The agreement was fairly good. (author)

  13. Experimental subcritical reactivity determinations employing APSD measurements with pulse mode detectors in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    This work aims to determine experimentally the subcritical reactivity levels of several configurations of the IPEN/MB-01 reactor in an approach based on the subcritical kinetic model developed by Gandini and Salvatores. The procedure employs the measurements of the APSD (Auto Power Spectral Density) using pulse mode detectors. The proposed approach is based only on measured quantities such as counting rates and the parameters arising from the least square approach of the APSD. Other difficult quantity such as detector efficiencies is not needed in the method. Several measurements of APSD were performed in varying degrees of sub-criticality (up to around -7000 pcm). The APSD data were least-square fitted to get the prompt decay mode (α). Beside the startup source, an external neutron sources of Am-Be was installed near the core in order to improve neutron count statistics. The final experimental results are of very good quality. The experiment shows clearly that the classical one point kinetic theory cannot describe the measured reactivity. MCNP Keff results were compared to the corresponding experimental results. The agreement was fairly good. (author)

  14. 239Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental Reactor Benchmarks

    Science.gov (United States)

    Peneliau, Y.; Litaize, O.; Archier, P.; De Saint Jean, C.

    2014-04-01

    A large set of nuclear data are investigated to improve the calculation predictions of the new neutron transport simulation codes. With the next generation of nuclear power plants (GEN IV projects), one expects to reduce the calculated uncertainties which are mainly coming from nuclear data and are still very important, before taking into account integral information in the adjustment process. In France, future nuclear power plant concepts will probably use MOX fuel, either in Sodium Fast Reactors or in Gas Cooled Fast Reactors. Consequently, the knowledge of 239Pu cross sections and other nuclear data is crucial issue in order to reduce these sources of uncertainty. The Prompt Fission Neutron Spectra (PFNS) for 239Pu are part of these relevant data (an IAEA working group is even dedicated to PFNS) and the work presented here deals with this particular topic. The main international data files (i.e. JEFF-3.1.1, ENDF/B-VII.0, JENDL-4.0, BRC-2009) have been considered and compared with two different spectra, coming from the works of Maslov and Kornilov respectively. The spectra are first compared by calculating their mathematical moments in order to characterize them. Then, a reference calculation using the whole JEFF-3.1.1 evaluation file is performed and compared with another calculation performed with a new evaluation file, in which the data block containing the fission spectra (MF=5, MT=18) is replaced by the investigated spectra (one for each evaluation). A set of benchmarks is used to analyze the effects of PFNS, covering criticality cases and mock-up cases in various neutron flux spectra (thermal, intermediate, and fast flux spectra). Data coming from many ICSBEP experiments are used (PU-SOL-THERM, PU-MET-FAST, PU-MET-INTER and PU-MET-MIXED) and French mock-up experiments are also investigated (EOLE for thermal neutron flux spectrum and MASURCA for fast neutron flux spectrum). This study shows that many experiments and neutron parameters are very sensitive to

  15. Analysis of the Pecore experimental programme carried out on Masurca and neutronics calculations for the PEC reactor

    International Nuclear Information System (INIS)

    A number of critical configurations installed in the Masurca experimental reactor of the French Commissariat a l'energie atomique (CEA) in Cadarache were used to simulate, by means of certain approximations, the neutron properties of the core of the PEC reactor. As the amount of plutonium available was insufficient, the Pecore experiment was carried out by using two different fuel zones: one with mixed U and Pu oxide, the other with enriched U oxide. However, an effort was made to keep the two zones similar as far as the neutron energy spectrum was concerned. The first interpretation of Pecore was carried out by the CEA with the assistance of staff of the Italian National Nuclear Energy Committee (CNEN), on the basis of French nuclear data and calculation methods. The paper describes the interpretation made in Italy on the basis of the codes and cross-sections used for the PEC neutron calculations, so that use could be made in this project of the discrepancies between calculation and experiment evaluated for Pecore. The characteristics compared were: critical mass, negative reactivity introduced by the control rods, ratios of reaction rates, efficiency of the nickel-base reflector and the axial and radial power curves. By way of a general conclusion, it can be stated that the interpretations of the CEA and the CNEN are in good agreement; the differences which they show are nearly always within the margin of experimental error. In addition to the main results of the Pecore experiment, the paper deals with certain aspects of: the PEC neutron calculation, from the definition of fuel enrichment to the calculation of the power level reached in each element, and from the updating of reactivity coefficients to the formulation of refuelling strategy. For some sectors, the results of the Pecore experiment had to be taken into account. (author)

  16. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  17. Experimental and calculational study of temperature distributions in deformed model fuel assemblies of fast reactors

    International Nuclear Information System (INIS)

    Experimental and calculational data tastify to absence of temperature nonuniformity stabilization in fuel assembly peripheral area. The effect of fuel lattice deformation on the fuel assembly temperature field at shroud crushing in the core centre is demonstrated. 17 refs.; 21 figs

  18. Failure mode and effect analysis experimental reliability determination for the CANDU reactor equipment

    International Nuclear Information System (INIS)

    This paper describes the experimental tests performed in order to prove the reliability parameters for certain equipment manufactured in INR Pitesti, for NPP Cernavoda. The tests were provided by Technical Specifications and test procedures. A comparison, referring to the reliability parameters, between Canadian equipment and INR manufactured equipment ones is also given. The results of tests and conclusions are shown. (author)

  19. Studies of thermal-reactor benchmark-data interpretation: experimental corrections

    International Nuclear Information System (INIS)

    Experimental values of integral parameters of the lattices studied in this report, i.e., the MIT(D2O) and TRX benchmark lattices have been re-examined and revised. The revisions correct several systematic errors that have been previously ignored or considered insignificant. These systematic errors are discussed in detail. The final corrected values are presented

  20. Experimental study of the supercritical water oxidation of recalcitrant compounds under hydrothermal flames using tubular reactors.

    Science.gov (United States)

    Cabeza, Pablo; Bermejo, M Dolores; Jiménez, Cristina; Cocero, M José

    2011-04-01

    The hydrothermal flame is a new method of combustion that takes place in supercritical water oxidation reactions when the temperature is higher than the autoignition temperature. In these conditions, waste can be completely mineralized in residence times of milliseconds without the formation of by-products typical of conventional combustion. The object of this work is to study the hydrothermal flame formation in aqueous streams with high concentrations of recalcitrant compounds: an industrial waste with a high concentration of acetic acid and various concentrated solutions of ammonia. A tubular reactor with a residence time of 0.7 s was used. Oxygen was used as the oxidant and isopropyl alcohol (IPA) as co-fuel to reach the operation temperature required. The increase of IPA concentrations in the feeds resulted in a better TOC removal. For mixtures containing acetic acid, 99% elimination of TOC was achieved at temperatures higher than 750 °C. In the case of mixtures containing ammonia, TOC removals reached 99% while maximum total nitrogen removals were never higher than 94%, even for reaction temperatures higher than 710 °C. Ignition was observed at concentrations as high as 6% wt NH(3) with 2% wt IPA while at IPA concentrations below 2% wt IPA, the ammonia did not ignite.

  1. 14MeV neutron irradiation experiment on window materials for fusion experimental reactor

    International Nuclear Information System (INIS)

    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 106-1011n/cm2/sec. The photon emission efficiency of F-center luminescence of the sapphire was 2200 ± 700photons/MeV, while the efficiency of F+-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)

  2. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  3. Scientific and technical conference Thermophysical experimental and calculating and theoretical studies to justify characteristics and safety of fast reactors. Thermophysics-2012. Book of abstracts

    International Nuclear Information System (INIS)

    The collection includes abstracts of reports of scientific and technical conference Thermophysics-2012 which has taken place on October 24-26, 2012 in Obninsk. In abstracts the following questions are considered: experimental and calculating and theoretical studies of thermal hydraulics of liquid-metal cooled fast reactors to justify their characteristics and safety; physico-chemical processes in the systems with liquid-metal coolants (LMC); physico-chemical characteristics and thermophysical properties of LMC; development of models, computational methods and calculational codes for simulating processes of of hydrodynamics, heat and mass transfer, including impurities mass transfer in the systems with LMC; methods and means for control of composition and condition of LMC in fast reactor circuits on impurities and purification from them; apparatuses, equipment and technological processes at the work with LMC taking into account the ecology, including fast reactors decommissioning; measuring techniques, sensors and devices for experimental studies of heat and mass transfer in the systems with LMC

  4. Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

    International Nuclear Information System (INIS)

    The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

  5. Geometrical modification transfer between specific meshes of each coupled physical codes. Application to the Jules Horowitz research reactor experimental devices

    International Nuclear Information System (INIS)

    The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author)

  6. Overview on stability of natural-circulation-cooled boiling water reactors during start-up. An experimental and modeling analysis

    International Nuclear Information System (INIS)

    This paper provides an overview on numerical and experimental work focused on flashing-induced instabilities. These instabilities may occur in natural circulation two-phase systems when operated at low pressure and low power. Therefore they are of special interest for the start-up phase of natural circulation Boiling Water Reactors. The work presented in this paper has been performed within the framework of the NACUSP project (European-Union Fifth Framework Program). Experiments were carried out on a steam/water natural circulation loop (CIRCUS), built at the Delft University of Technology. Information was gained on the characteristics of the flow oscillations and on the void fraction production during flashing in stationary and transient conditions. A 3-D flow-pattern visualization was achieved by means of advanced instrumentation, namely wire-mesh sensors. On the basis of the experimental results, an assessment of existing drift-flux models was performed for flashing flow. The most suitable drift-flux model was implemented in the 4-equations two-phase model FLOCAL, developed at the Forschungszentrum Rossendorf (FZR, Germany). The model allows for the liquid and steam to be in thermal non-equilibrium and, via drift-flux models, to have different velocities. A detail comparison between simulations and experiments is reported. (author)

  7. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    International Nuclear Information System (INIS)

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  8. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang-Soon; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik; Kim, Hwan-Yeol; Kim, Hee-dong

    2005-04-01

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  9. 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.

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

    International Nuclear Information System (INIS)

    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

  11. Experimental demonstration of the reverse flow catalytic membrane reactor concept for energy efficient syngas production. Part 1: Influence of operating conditions

    NARCIS (Netherlands)

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

    2007-01-01

    In this contribution the technical feasibility of the reverse flow catalytic membrane reactor (RFCMR) concept with porous membranes for energy efficient syngas production is investigated. In earlier work an experimental proof of principle was already provided [Smit, J., Bekink, G.J., van Sint Annala

  12. The ability to create NTD silicon technology in the IRT-T reactor in a horizontal experimental channel with one-side access

    Science.gov (United States)

    Varlachev, V. A.; Golovatsky, A. V.; Emets, E. G.; Butko, Ya A.

    2016-06-01

    The article shows the ability of creation of neutron transmutation doping (NTD) of monocrystalline silicon technology in the reactor's channel, which has a one-side access. In the article a distribution of thermal neutron flux through the length of channel and it's radius, neutron spectrum were obtained which confirmed that horizontal experimental channel HEC-1 is suitable for NTD.

  13. Theoretical and experimental substantiation of the fusion reactor divertor and first wall shield concept by means of lithium capillary porous systems

    International Nuclear Information System (INIS)

    Basic properties of the lithium capillary porous systems (LCPS) for protection of divertor and the first wall shield TNR are described. The LCPS behaviour under conditions of impact of high stationary and pulse heat loads are studied. Lithium ions and neutrons migration in the tokamak reactor with liquid lithium is studied numerically and lithium flow in the crossed magnetic field experimentally. The LCPS behaviour is studied also experimentally under the conditions of the tokamak plasma impact

  14. Experimental analysis of a nuclear reactor prestressed concrete pressure vessels model

    International Nuclear Information System (INIS)

    A comprehensible analysis was made of the performance of each set of sensors used to measure the strain and displacement of a 1/20 scale Prestressed Concrete Pressure Vessel (PCPV) model tested at the Instituto de Pesquisas Energeticas e Nucleares (IPEN). Among the three Kinds of sensors used (strain gage, displacement transducers and load cells) the displacement transducers showed the best behavior. The displacemente transducers data was statistically analysed and a linear behavior of the model was observed during the first pressurizations tests. By means of a linear statistical correlation between experimental and expected theoretical data it was found that the model looses the linearity at a pressure between 110-125 atm. (Author)

  15. Removal of carbon dioxide in an experimental powder-particle spouted bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Haghnegahdar, M.R.; Hatamipour, M.S.; Rahimi, A. [University of Isfahan, Esfahan (Iran). Dept. of Chemical Engineering

    2010-05-11

    The performance of a powder-particle spouted bed (PPSB) on the removal of CO{sub 2} is investigated. A laboratory scale PPSB is employed to investigate the effects of operating parameters such as approach to saturation temperature, static bed height, Ca/C molar ratio, inlet CO{sub 2} concentration and type of sorbent on CO{sub 2} removal efficiency. The experimental results show that the CO{sub 2} removal efficiency increases by increasing the static bed height, Ca/C molar ratio and inlet CO{sub 2} concentration, and decreases by increasing the approach to saturation temperature and superficial gas velocity. Also it is concluded that maximum CO{sub 2} removal efficiency could be up to 50% when approach to saturation temperature is 8K, Ca/C molar ratio is 1.4 and the static bed height is 0.225 m.

  16. 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.

  17. Theoretical and experimental studies of non-linear structural dynamics of fast breeder reactor fuel elements

    International Nuclear Information System (INIS)

    Descriptions are presented of theoretical and experimental studies of the deformation behaviour of fast-breeder fuel elements as a consequence of extreme impulsive stresses produced by an incident. The starting point for the studies is the assumption that local disturbances in a fuel element have resulted in a thermal interaction between fuel and sodium and in a corresponding increase in pressure. On the basis of the current state of knowledge, the possibility cannot be ruled out that this pressure build-up may lead to the bursting of the fuel-element wrapper, to the propagation of pressure in the core, and to coherent structural movements and deformations. A physical model is established for the calculation of the dynamic response of elastic-plastic beam systems, and the differential equations of p motion for the discrete equivalent system are derived with the aid of D'Alembert's principle. On this basis and with the aid of a semi-empirical pin-bundle model, an appropriate computer program allows a static and dynamic analysis to be obtained for a complete fuel element. In the experimental part of the study, a description is given of static and impulsive loading tests on 1:1 SNR-like fuel-element models. Making use of measured impact forces and of known material characteristics, it was possible to a large extent for the experiments to be reproduced by calculations. In agreement with existing experience from explosion experiments on 1:1 core models, the results (of relevance for fast-breeder safety and in particular the SNR-300) show that only local limited deformations occur and that the compact fuel-element and core structure constitutes an effective inherent barrier in the presence of extreme incident stresses. (author)

  18. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  19. Experimental study of lactose hydrolysis and separation in cstr-uf membrane reactor

    Directory of Open Access Journals (Sweden)

    M. Namvar-Mahboub

    2012-09-01

    Full Text Available In this study, the effect of processing conditions on the performance of continuous stirred tank -ultrafiltration (CSTR-UF in dead - end mode was investigated. An UF membrane with a molecular weight cutoff of 3 kDa made of regenerated cellulose material was used to separate enzyme from products. The effect of operating pressure ranging between 2 and 5 bar and time on the performance of the CSTR-UF system was studied. The experiments were performed with a 0.139 molar aqueous solution of lactose as feed. According to the experimental data, the lactose concentration in the permeate decreased with time due to concentration polarization and hydrolysis. It was found that the rejection factor of lactose increases from 33 to 77% with time from 5 to 85 min. Permeation flux of the membrane was evaluated in terms of pure water flux (PWF and lactose aqueous solution. Results showed that a high operating pressure led to a high permeation flux for both mentioned cases. Also, adding lactose and enzyme to pure water caused a reduction of the permeation flux due to concentration polarization.

  20. 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)

  1. 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.

  2. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  3. Experimental and numerical investigation of sub-cooled boiling, condensation, and void flashing in nuclear heating reactor test loop

    International Nuclear Information System (INIS)

    This paper describes experimental and numerical investigations of sub-cooled boiling, condensation, and void flashing in the HRTL-5 test loop, which simulates the primary loop of a 5 MW nuclear heating reactor. A drift-flow model of two-phase with four governing equations was used, in which sub-cooled boiling, condensation, and void flashing have been taken into account. Based on the mathematical model, a program has been developed for analyzing the natural circulation system. As parameters, inlet sub-cooling, system pressure, and heat flux are varied. For comparison, some simplified models, which are designed to reveal the importance of sub-cooled boiling, condensation, flashing in the HRTL-5 test loop, are adopted in the program. The results show: (1) subcooled boiling, condensation, and void flashing may have great influence on the distribution of the void fraction and more intense at low system pressure; (2) the calculation of them is correlative and interactive other than independent; (3) for a system with short heated section, long riser, and low pressure, it is possible to reach 'boiling out of the core', where there is almost no void in the heated section, but much in the riser. (orig.)

  4. 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)

  5. Calculations of energy losses due to atomic processes in tokamaks with applications to the International Thermonuclear Experimental Reactor divertor

    International Nuclear Information System (INIS)

    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) [Rebut et al., Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, in press)]. 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 depend upon 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 quasianalytic 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 from the scrape-off layer, and divertor plasma and impurity radiation losses from the divertor plasma. This analysis indicates that bremsstrahlung from the plasma center and impurity radiation from the plasma edge and divertor plasma can each play a significant role in reducing the power to the divertor plates, and identifies many of the factors which determine the relative role of each process. For instance, for radiation losses in the divertor to be large enough to radiate the power in the divertor for high power experiments, a neutral fraction of 10-3 to 10-2 and an impurity recycling rate of neτrecycle of ∼1016 s m-3 will be required in the divertor

  6. Preliminary analysis in support to the experimental activities on the mixing process in the pressurizer of a small modular reactor integrated primary system

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Samira R.V.; Lira, Carlos A.B.O.; Bezerra, Jair L.; Silva, Mario A.B.; Silva, Willdauany C.F., E-mail: samiraruana@gmail.com [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Lima, Fernando R.A., E-mail: falima@crcn.gov.br [Centro Regional de Ciencias Nucleares (CRCN/CNEN-NE), Recife, PE (Brazil); Otero, Maria E.M.; Hernandez, Carlos R.G., E-mail: mmontesi@instec.cu [Department of Nuclear Engineering, InSTEC/CUBA, Higher Institute of Technology and Applied Science, La Habana (Cuba)

    2015-07-01

    Nowadays, there is a renewed interest in the development of advanced/innovative small and medium sized modular reactors (SMRs). The SMRs are variants of the Generation IV systems and usually have attractive characteristics of simplicity, enhanced safety and require limited financial resources. The concept of the integrated primary system reactor (IPSR) is characterized by the inclusion of the entire primary system within a single pressure vessel, including the steam generator and pressurizer. The pressurizer is located within the reactor vessel top, this configuration involves changes on the techniques and is necessary investigate the boron mixing. The present work represents a contribution to the design of an experimental facility planned to provide data relevant for the mixing phenomena in the pressurizer of a compact modular reactor. In particular, in order to evaluate the boron concentration in the surge orifices to simulate the in-surge and out-surge in a facility, scaled 1:200, respect to the ¼ of the pressurizer. The facility behavior studied from one inlet and one outlet of the test section with represent one in-surge e one out-surge the pressurizer of a small modular reactor integrated primary system. (author)

  7. Preliminary analysis in support to the experimental activities on the mixing process in the pressurizer of a small modular reactor integrated primary system

    International Nuclear Information System (INIS)

    Nowadays, there is a renewed interest in the development of advanced/innovative small and medium sized modular reactors (SMRs). The SMRs are variants of the Generation IV systems and usually have attractive characteristics of simplicity, enhanced safety and require limited financial resources. The concept of the integrated primary system reactor (IPSR) is characterized by the inclusion of the entire primary system within a single pressure vessel, including the steam generator and pressurizer. The pressurizer is located within the reactor vessel top, this configuration involves changes on the techniques and is necessary investigate the boron mixing. The present work represents a contribution to the design of an experimental facility planned to provide data relevant for the mixing phenomena in the pressurizer of a compact modular reactor. In particular, in order to evaluate the boron concentration in the surge orifices to simulate the in-surge and out-surge in a facility, scaled 1:200, respect to the ¼ of the pressurizer. The facility behavior studied from one inlet and one outlet of the test section with represent one in-surge e one out-surge the pressurizer of a small modular reactor integrated primary system. (author)

  8. Experimental power reactor

    International Nuclear Information System (INIS)

    The following five topics are discussed using figures and diagrams: (1) energy storage and transfer program, (2) thermomechanical analysis, (3) a steam dual-cycle power conversion system for the EPR, (4) EPR tritium facility scoping studies, and (5) vacuum systems

  9. Thermal study of a non adiabatic differential calorimeter used for nuclear heating measurements inside an experimental channel of the Jules Horowitz Reactor

    Science.gov (United States)

    Reynard-Carette, C.; Lyoussi, A.; Brun, J.; Muraglia, M.; Carette, M.; Janulyte, A.; Zerega, Y.; André, J.; Bignan, G.; Chauvin, J.-P.; Fourmentel, D.; Gonnier, C.; Guimbal, P.; Malo, J.-Y.; Villard, J.-F.

    2012-11-01

    New online in-pile measurement methods are crucial during irradiations in Material Testing Reactors (MTR) for a better understanding of accelerated material ageing and nuclear fuel behaviour. In particular, instrumentation for measurements of one relevant parameter: nuclear heat deposition rate, called nuclear heating, has to be improved. The knowledge of this quantity is a great interest for various safety, scientist and end-user requirements (design of specific irradiation devices and associated cooling systems with imposed conditions). This paper focuses on thermal experimental and numerical studies carried out under non irradiation conditions on an in-pile calorimeter dedicated to nuclear heating quantification inside a new experimental device which will be dedicated to the experimental condition mapping (neutron and photon fluxes and nuclear heating) inside the JHR experimental channels. Experimental results concerning the calorimeter response during its electrical calibration (calorimeter design is compared to a single calorimeter which gives promising results (miniaturization, higher sensitivity).

  10. Design, implementation and cost-benefit analysis of a dynamic testing program in the Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    Dynamic tests have been performed for many years in commercial pressurized and boiling water reactors. The purpose of this study was to evaluate the technological and economical feasibility of extending the current light water reactor testing procedures to both present and future liquid metal fast breeder reactors. A 38 node linearized, lumped parameter, EBR-II system model was developed. This model was analyzed to obtain the predicted system time and frequency response for reactivity perturbations, intermediate heat exchanger secondary inlet sodium temperature perturbation frequency response, and various system nodal frequency response sensitivities

  11. 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)

  12. Study on uranium-water multiplicative means of the (RESUCO-Subcritical experimental reactor of uranium with oxygen) subcritical assembly by pulsed neutron technique

    International Nuclear Information System (INIS)

    The effective multiplication factor and the nuclear parameters associated with the variation of (RESUCO- Subcritical Experimental Reactor of Uranium with Oxygen) Subcritical Assembly Configuration, using pulsed neutron technique are analysed. BF3 detectors were used to detect the variation of thermal neutrons in the system, positioned parallelly to fuel elements, and a proton recoil detector was used for monitoring the neutron generation. (M.C.K.)

  13. 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

    International Nuclear Information System (INIS)

    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)

  14. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    International Nuclear Information System (INIS)

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations

  15. 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.

  16. Experimental investigations on carbonation of sodium aerosol generated from sodium fire in the context of fast reactor safety

    International Nuclear Information System (INIS)

    Highlights: • Investigated carbonation of aerosol released from sodium fire. • Investigated influence of % RH and CO2 content in air on carbonation process. • Deployed a novel technique for assay of different specimens in Na-aerosol. • This work can be used to predict chemical assay of aerosols in large sodium fire. - Abstract: Carbonation of sodium aerosols is the most important aspects to be considered for the evaluation of chemical hazards as a part of fast reactor safety studies. The sodium oxide, immediately formed as the combustion product due to sodium fire, undergoes chemical changes to NaOH, Na2CO3 and NaHCO3 upon reactions with moisture and CO2 prevailed in the atmosphere. Of which, hydroxide aerosols are highly corrosive and harmful, and it has stringent concentration limit for human exposure. Hence, in order to assess the condition for human intervention in the event of sodium fire, chemical composition of aerosols resulting from controlled sodium fires in a closed Aerosol Test Facility was investigated. The real time chemical species of aerosols generated from sodium fire and the effect of relative humidity (RH) and carbon dioxide concentration in air on carbonation have been studied. The experiments were carried out with the initial mass concentration of ∼4 g m−3, RH between 20% and 90% and the CO2 concentration in surrounding environment at 390 and 280 ppm. It is observed from the experimental study that aerosols are enriched with NaOH (0.8 mol fraction) in the beginning stage (samples collected during first few minutes after sodium fire) when surrounding atmosphere contains any of the following compositions – (i) ∼90% RH and 390 ppm CO2, (ii) ∼90% RH and 280 ppm CO2 or (iii) 50% RH and 280 ppm CO2 whereas they are almost equally distributed between NaOH and Na2CO3 in the beginning stage when the atmosphere has any of the compositions (i) 50% RH, 390 ppm CO2, (ii) 20% RH, 390 ppm CO2 or (iii) 20% RH, 280 ppm CO2. Carbonation of

  17. Experimentation and simulation of tin oxide deposition on glass based on the SnCl4 hydrolysis in an in-line atmospheric pressure chemical vapor deposition reactor

    International Nuclear Information System (INIS)

    Tin oxide thin films were deposited on glass substrates in an in-line atmospheric pressure chemical vapor deposition reactor under various conditions, which were numerically simulated using a commercial package. A soda-lime glass sheet was used as a deposition substrate, and SnCl4 and H2O in gas phase were separately supplied as the precursor and the oxidizer, respectively. By assuming that the main chemical reactions followed the Rideal–Eley mechanism, the experimentally determined deposition rates were fitted to obtain the reaction factors needed to describe the deposition process. The role of barrier gas injection for minimizing unwanted surface reaction or particle generation inside of the reactor, and not on the target (glass backplane itself) has been elucidated. Furthermore, the optimal operating conditions for the deposition on glass with the additives such as CH3OH and HF have been investigated. - Highlights: • Tin oxide deposition on glass based on the SnCl4 hydrolysis in an in-line reactor • Simulations using Rideal–Eley mechanism were compared to experimental observations. • Results: activation energy—79.3 kJ/mol and frequency factor—1.93 × 1010 m4/kmol·s • Influences of additives on transmittances, hazes, and electrical resistivities

  18. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  19. Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II

    International Nuclear Information System (INIS)

    This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings

  20. Experimental study of the neutronics of the first gas cooled fast reactor benchmark assembly (GCFR phase I assembly)

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, S.K.

    1976-12-01

    The Gas Cooled Fast Reactor (GCFR) Phase I Assembly is the first in a series of ZPR-9 critical assemblies designed to provide a reference set of reactor physics measurements in support of the 300 MW(e) GCFR Demonstration Plant designed by General Atomic Company. The Phase I Assembly was the first complete mockup of a GCFR core ever built. A set of basic reactor physics measurements were performed in the assembly to characterize the neutronics of the assembly and assess the impact of the neutron streaming on the various integral parameters. The analysis of the experiments was carried out using ENDF/B-IV based data and two-dimensional diffusion theory methods. The Benoist method of using directional diffusion coefficients was used to treat the anisotropic effects of neutron streaming within the framework of diffusion theory. Calculated predictions of most integral parameters in the GCFR showed the same kinds of agreements with experiment as in earlier LMFBR assemblies.

  1. Analysis of a possible experimental assessment of a prototype fuel element containing burnable poison in the RA-3 reactor

    International Nuclear Information System (INIS)

    The Argentine RA-3 research reactor (5 MW) is presently operated with LEU fuel by the National Atomic Energy Commission (CNEA). It belongs to the group of nuclear installations controlled, from the radiological and nuclear safety point of view, by the Nuclear Regulatory Authority (ARN). A new type of fuel elements containing burnable absorbers, with similar enrichment as the standard fuel elements but greater fissile contents, has recently been proposed for a new Argentine reactor design (RRR). In this framework the ARN considers interesting, if technically possible, the performance of an experiment in the RA-3 reactor. The experiment might enable, for such fuel element containing burnable poison, the verification of its neutronic behaviour under irradiation as well as a validation of the calculation line by comparison to measured values. It should be desirable that such experiment could reproduce as much as possible those conditions estimated for the RRR reactor, still under design in Argentina, having Silicide fuel elements with burnable poison, in the shape of cadmium wires in their structure. We here analyse a possible experiment consisting in the loading of a prototype fuel element with burnable poison in a normally loaded RA-3 core configuration. It would essentially be a standard RA-3 fuel element, having cadmium wires in its frame. This experiment would enable the verification of the prototype behaviour under irradiation, its operation limits and conditions, and particularly, the reactivity safety margins established in Argentine Standards, both calculated and measured. The main part of the experiment would imply some 200 full power days of operation at 5 MW, which would be drastically reduced if the reactor power is increased to 10 MW, as foreseen. We also show that under the proposed conditions, the experiment would not represent a significant penalty to the reactor normal operation. (author)

  2. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.

    2014-05-01

    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  3. Experimental determination of prompt neutron decay constant of IPEN/MB-01 reactor by Rossi-{alpha} method; Determinacao experimental da constante de decaimento de neutrons prontos do reator IPEN/MB-01 via o metodo Rossi-{alpha}

    Energy Technology Data Exchange (ETDEWEB)

    Kuramoto, Renato; Santos, Adimir dos; Jerez, Rogerio; Bitelli, Ulysses D' Utra; Diniz, Ricardo; Madi, Tufic [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: ryrkuram@ipen.br

    2005-07-01

    One major objective of this work is to experimentally estimate the prompt neutron decay constant, {alpha}, on the IPEN/MB-01 research reactor at Sao Paulo. In order to achieve our goal, we will use a microscopic noise technique called Rossi-{alpha} method. This method is based on the statistical nature of the fission-chain process. Using a coincidence acquisition system, the rationale is to experimentally determine the probability distribution of detecting neutrons from the same chain. Through a leastsquares fit of this distribution we estimate {alpha}=({beta}{sub eff}-{rho})/{lambda}. The {alpha} parameter will be measured for different sub-critical levels, and the ratio {beta}{sub eff}/{lambda} is obtained via extrapolation to {rho}=0. A specific acquisition system for Rossi-{alpha} measurements has been developed in order to achieve our objective. This system is based on a multichannel scaler controlled by virtual instruments that records the timing of all neutron events, allowing data analysis during the acquisition. The first measurements were performed using one BF{sub 3} detectors positioned at the center of the core of IPEN/MB-01 research reactor. In this work we will present a preliminary set of Rossi-{alpha} measurements performed on the IPEN/MB-01 research reactor at Sao Paulo. (author)

  4. China Experimental Fast Reactor(CEFR)——Criterion of Criticality for Reactor With External Neutron Source

    Institute of Scientific and Technical Information of China (English)

    ZHAOYu-sen

    2003-01-01

    There is a neutron source with 109 s-1 neutrons in core of CEFR during start up test and operation of CEFR. For judging the criticality of reactor with external neutron source and near criticality, it is important that the neutron level changes in core with time must be understood after introducing positive reactivity to core with external neutron source.

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

    NARCIS (Netherlands)

    Patil, Charudatta S.; Sint Annaland, van Martin; 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

  6. 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

  7. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  8. Experimental investigations for determination of heat-transfer coefficients and temperature fields in simulated fuel assemblies of BREST reactor with fuel elements spaced by transverse grids

    International Nuclear Information System (INIS)

    The consideration is given to heat transfer and temperature fields in fuel pin bundles with transverse spacer grids (s/d =1.33) equally spaced along energy deposition length. Experimental data are obtained on two simulated 37-rod core assemblies: one assembly is with uniform geometry along the cross-section and in the other there is nonheated rod simulating supporting pipe in fuel assembly of reactor with heavy coolant. Eutectic Na-K alloy is used as coolant. Nusselt numbers and temperature nonuniformity along the perimeter of measurement fuel element simulator obtained in these assemblies are compared as well as available data for finned (wire to wire) fuel rods

  9. Experimental investigation into the stress behavior of a reactor coolant pump impeller subjected to a wide range of flow conditions

    International Nuclear Information System (INIS)

    The instrumentation scheme used to successfully monitor static and dynamic strains at the vanes of a reactor coolant pump impeller during simulated reactor flow testing is described. Extensive bench tests were performed to optimize the placement of weldable strain gauges near the tips of the vanes, where an extremely hostile flow environment existed. Special instrument cable routing techniques plus telemetry equipment were employed to record the strain gauge time history records for on line analysis, as well as for subsequent, advanced digital signal processing. Typical strain data are presented which, in general terms, describe the steady state (static) and the dynamic components of the strain field at the pump vanes as a function of flow. (author)

  10. Aqueous two-phase micellar systems in an oscillatory flow micro-reactor: Study of perspectives and experimental performance

    OpenAIRE

    A. M. LOPES; Silva, Daniel Pereira da; A.A. Vicente; Pessoa Júnior, Adalberto; Teixeira, J. A.

    2011-01-01

    Aqueous two-phase micellar systems (ATPMS) are micellar surfactant solutions with physical properties that make them very efficient for the extraction/concentration of biological products. In this work the main proposal that has been discussed is the possible applicability and importance of a novel oscillatory flow micro-reactor (micro-OFR) envisaged for parallel screening and/or development of industrial bioprocesses in ATPMS. Based on the technology of oscillatory flow mixing (OFM), this ba...

  11. Idaho National Laboratory Experimental Program to Measure the Flow Phenomena in a Scaled Model of a Prismatic Gas-Cooled Reactor Lower Plenum for Validation of CFD Codes

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-09-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a prismatic gas-cooled reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A description of the scaling analysis, experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that will be presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic gas-cooled reactor (GCR) similar to a General Atomics Gas-Turbine-Modular Helium Reactor (GTMHR) design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. The flow in the lower plenum consists of multiple jets injected into a confined cross flow - with obstructions. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to approximate flow scaled to that expected from the staggered parallel rows of posts in the reactor design. The model is fabricated from clear, fused quartz to match the refractive-index of the mineral oil working fluid. The benefit of the MIR technique is that it permits high-quality measurements to be obtained without locating intrusive transducers that disturb the flow field and without distortion of the optical paths. An advantage of the INL MIR system is its large size which allows improved spatial and temporal resolution compared to similar facilities at smaller scales. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements

  12. Activation analysis and waste management for blanket materials of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

    International Nuclear Information System (INIS)

    The preliminary studies of the activation analysis and waste management for blanket materials of the multi-functional experimental fusion–fission hybrid reactor, i.e. Multi-Functional eXperimental Fusion Driven Subcritical system named FDS-MFX, were performed. The neutron flux of the FDS-MFX blanket was calculated using VisualBUS code and Hybrid Evaluated Nuclear Data Library (HENDL) developed by FDS Team. Based on these calculated neutron fluxes, the activation properties of blanket materials were analyzed by the induced radioactivity, the decay heat and the contact dose rate for different regions of the FDS-MFX blanket. The safety and environment assessment of fusion power (SEAFP) strategy, which was developed in Europe, was applied to FDS-MFX blanket for the management of activated materials. Accordingly, the classification and management strategy of activated materials after different cooling time were proposed for FDS-MFX blanket

  13. Early detection of leakage in reactor coolant system of nuclear plant using Kalman filter. Experimental verification and consideration on response time for detection under long-sustained operation

    International Nuclear Information System (INIS)

    This report concerns on the performance of leakage detection systems for the coolant of nuclear reactor. The detection by Kalman filter is based on the change of total mass of coolant. Two points are analyzed: (1) The performance of this method is experimentally investigated by a bench-scale coolant system model. (2) The response time to detect is investigated by simulation under the condition of actual operation. This means that leakage occurs later from the system starting. The experimental result shows that proposed system can detect 0.06% loss of coolant. Therefore this method can be considered realizable. And by simulation it is cleared that occurrence of leakage in any time can be detected within 14 minutes. Comparing with the design target of the present system, this system is satisfying. Through above discussion, this system is concluded to be applicable and useful to an actual plant. (author)

  14. The application of the multiple transient technique for the experimental determination of the relative abundances and decay constants of delayed neutrons of the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, Sao Paulo, SP 05508-000 (Brazil)]. E-mail: asantos@ipen.br; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, Sao Paulo, SP 05508-000 (Brazil); Jerez, Rogerio [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, Sao Paulo, SP 05508-000 (Brazil); Mai, Luiz Antonio [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, Sao Paulo, SP 05508-000 (Brazil); Yamaguchi, Mitsuo [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP, Av. Prof. Lineu Prestes, 2242, Cidade Universitaria, Sao Paulo, SP 05508-000 (Brazil)

    2006-07-15

    An in-pile experiment for the determination of the relative abundances and decay constants of delayed neutrons has been successfully performed at the IPEN/MB-01 research reactor facility. The experimental data are of good quality and can be used to validate theoretical predictions of the delayed neutron group constants based on the current knowledge of the fission products yields and emission probabilities for known precursors. The theory/experiment comparison shows that the current release of ENDF/B-VI, namely release 8, shows severe discrepancies in both relative abundances and in its first decay constant. The version revised at LANL shows very good progress in both aspects. JENDL3.3 shows the best performance in the C/E comparison. One of the main achievements of the experiment was the consistency of the measured first decay constant to that of {sup 87}Br. Finally, it is also shown preliminary experimental results for an eight-group model.

  15. Experimental study on buoyancy-driven exchange flows through breaches of a tokamak vacuum vessel in a fusion reactor under the loss-of-vacuum-event conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Kazuyuki; Tomoaki, Kunugi; Ogawa, Masurou; Seki, Yasushi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    1997-02-01

    As one of thermofluid safety studies in the International Thermonuclear Experimental Reactor, buoyancy-driven exchange flow behavior through breaches of a vacuum vessel (VV) has been investigated quantitatively by using a preliminary loss-of-vacuum-event (LOVA) apparatus that simulated the tokamak VV of a fusion reactor with a small-scaled model. To carry out the present experiments under the atmospheric pressure condition, helium gas and air were provided as the working fluids. The inside of the VV was initially filled with helium gas and the outside was atmosphere. The breaches on the VV under the LOVA condition were simulated by opening six simulated breaches to which were set the different positions on the VV. When the buoyancy-driven exchange flow through the breach occurred, helium gas went out from the inside of the VV through the breach to the outside and air flowed into the inside of the VV through the breach from the outside. The exchange rate in the VV between helium gas and air was calculated from the measured weight change of the VV with time since the experiment has started. experimental parameters were breach position, breach number, breach length, breach size, and breach combination. The present study clarifies that the relation between the exchange rate and the breach position of the VV depended on the magnitude of the potential energy from the ground level to the breach position, and then, the exchange rate decreased as the breach length increased and as the breach size decreased.

  16. Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant

    Science.gov (United States)

    Belyaev, I. A.; Genin, L. G.; Krylov, S. G.; Novikov, A. O.; Razuvanov, N. G.; Sviridov, V. G.

    2015-09-01

    The aim of this experimental investigation is to obtain information on the temperature fields and heat transfer coefficients during flow of liquid-metal coolant in models simulating an elementary cell in the core of a liquid heavy metal cooled fast-neutron reactor. Two design versions for spacing fuel rods in the reactor core were considered. In the first version, the fuel rods were spaced apart from one another using helical wire wound on the fuel rod external surface, and in the second version spacer grids were used for the same purpose. The experiments were carried out on the mercury loop available at the Moscow Power Engineering Institute National Research University's Chair of Engineering Thermal Physics. Two experimental sections simulating an elementary cell for each of the fuel rod spacing versions were fabricated. The temperature fields were investigated using a dedicated hinged probe that allows temperature to be measured at any point of the studied channel cross section. The heat-transfer coefficients were determined using the wall temperature values obtained at the moment when the probe thermocouple tail end touched the channel wall. Such method of determining the wall temperature makes it possible to alleviate errors that are unavoidable in case of measuring the wall temperature using thermocouples placed in slots milled in the wall. In carrying out the experiments, an automated system of scientific research was applied, which allows a large body of data to be obtained within a short period of time. The experimental investigations in the first test section were carried out at Re = 8700, and in the second one, at five values of Reynolds number. Information about temperature fields was obtained by statistically processing the array of sampled probe thermocouple indications at 300 points in the experimental channel cross section. Reach material has been obtained for verifying the codes used for calculating velocity and temperature fields in channels with

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

    Science.gov (United States)

    Kuşçu, Özlem Selçuk; Sponza, Delia Teresa

    2011-03-15

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

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

    International Nuclear Information System (INIS)

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

  19. Speech by Prime Minister Francois Fillon. Visit of the Jules Horowitz experimental reactor works on the Commissariat a l'Energie et aux Energies Alternatives site. Cadarache, May 3, 2010

    International Nuclear Information System (INIS)

    In this speech, the French Prime Minister evokes the present context, the importance of strategic technologies, and the challenge of investing in these technologies within a context of reduction of public expenses. He comments the decision of his government to finance research and education activities in different domains, and more specifically in the energy sector with this fourth generation Jules Horowitz experimental reactor. He recalls that the nuclear sector has always been very important to the eyes of the successive French governments, and outlines how this reactor will contribute to reactor operational optimization, lifetime extension and safety, nuclear fuel development, etc.

  20. Experimental investigation on plasma parameter profiles on a wafer level with reactor gap lengths in an inductively coupled plasma

    International Nuclear Information System (INIS)

    The gap length effect on plasma parameters is investigated in a planar type inductively coupled plasma at various conditions. The spatial profiles of ion densities and the electron temperatures on the wafer level are measured with a 2D probe array based on the floating harmonic method. At low pressures, the spatial profiles of the plasma parameters rarely changed by various gap lengths, which indicates that nonlocal kinetics are dominant at low pressures. However, at relatively high pressures, the spatial profiles of the plasma parameter changed dramatically. These plasma distribution profile characteristics should be considered for plasma reactor design and processing setup, and can be explained by the diffusion of charged particles and the local kinetics

  1. 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...... 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...... continuous aeration, could remove more than 5.5 g N/m2/day (at loads up to 8 g N/m2/day) by controlled variation of sequential aeration regimes. Daily averaged ratios of the surficial loads of O2 (oxygen) to NH4+ (ammonium) (LO2/LNH4) were close to 1.73 at this optimum. Real-time quantitative PCR based on 16...

  2. 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.

  3. Using the Bayes theorem of conditional probabilities to obtain the neutron u x of the RECH-1 experimental nuclear reactor at CCHEN

    International Nuclear Information System (INIS)

    The Bayes' theorem of conditional probabilities is shown as an alternative to unfold the neutron u x energy distribution, Φ (E), from neutron activation measurements of different target material, A∞i at RECH-1 reactor core at CCHEN and the corresponding tabulated cross sectionsσ, since A∞ i = J∞ o σi (E)Φ(E)d E,→A∞i = Σn j = 1σijΦj. The unfolding algorithms used in the analysis of target activation to obtain the neutron flux energy spectrum of a nuclear reactor core in similar devices worldwide, need an a priori knowledge of the spectrum shape to solve the inverse problem of obtaining Φj. The advantage of using the Bayes based unfolding iterative algorithm is that an a priori knowledge of the neutron u x spectrum is not needed to obtain the experimental spectrum. In fact it is possible to start iterating with an uniform distribution. This unfolding algorithm have been used recently to solve analogue problems in nuclear physics such obtaining feeding probabilities in total gamma spectroscopy beta decay measurements, and to obtain the neutron energy background at Can franc Underground Laboratories using 3He detectors embedded in individual polyethylene blocks of different sizes. Furthermore, using this algorithm to obtain neutron energy spectrum from a nuclear reactor is a novel work, extending the applicability of the method. In this work, due the activation measurements have not been done yet, an analysis of the unfolding algorithm is presented using real tabulated cross sections and starting the iterations from an uniform u x distribution

  4. Experimental study for research and development of a super fast reactor. (2) Oscillatory condensation of high temperature vapor directly discharged into sub-cooled liquid pool

    International Nuclear Information System (INIS)

    The measurement of pressure oscillation and the observation of condensation behavior of a vapor discharge into sub-cooled liquid cool has been carried out to obtain a basic data for the evaluation of safety of the LOCA in the supercritical pressure light water cooled fast reactor (Super Fast Reactor). In the experiment, HCFC 123 is used as the test fluid. HCFC 123 is easy for handling due to its low critical pressure and temperature, and therefore, the experimental conditions can be set easily to make systematic data. The vapor at high temperature is discharged into the sub-cooled liquid pool through a submerged single pipe vertically fixed. The oscillatory condensation is observed. The condensation oscillation produces pressure oscillation in the liquid pool. The condensing interface area becomes small as the increase of the degree of sub-cooling. The pressure frequency has a period of millisecond order and the frequency and amplitude of the pressure oscillation increase with increasing the degree of sub-cooling and mass flux of the vapor, like the results of some conventional water vapor injection tests. In the present study, it is also consistently discussed the influence of the vapor temperature, mass flux, mass flow rate, back pressure of the liquid pool, pipe diameter and the degree of sub-cooling on the pressure amplitude and condensation behavior. (author)

  5. Some implications of radiation-induced property changes in austenitic stainless steels on ITER [International Thermonuclear Experimental Reactor] first-wall design and performance

    International Nuclear Information System (INIS)

    New data on radiation-induced hardening, low-temperature creep and potential susceptibility (sensitization) to aqueous corrosion have been obtained on various heats of austenitic stainless steel (including type 316) irradiated at 60--400 degree C to 7--13 dpa. The data were obtained from spectral-tailoring reactor experiments, whose radiation-damage parameters are similar to those in the proposed International Thermonuclear Experimental Reactor (ITER) first-wall (FW) and blanket design. Austenitic stainless steels were found to increase significantly in strength at 60--330 degree C, to have higher irradiation-creep rates at 60 degree C than at 200--400 degree C, and to show radiation-induced changes in electrochemical properties at 200--400 degree C. These data on several radiation-induced property changes suggest that type 316 steel may be an adequate material for the FW of ITER. However, there is definitely a need for new data on fracture-toughness and on fatigue behavior below 400 degree C, as well as more data on irradiation-creep and effects of irradiation on corrosion properties to better define temperature and dose dependencies for more detailed design analyses. Cold-working should remain an optional as-fabricated condition for the FW of ITER. Many properties of SA and CW 316 become similar after irradiation at 60--400 degree C. The higher initial yield-strength of CW 316 will allow higher design stress and elastic strain limits. 31 refs., 10 figs

  6. Experimental study on application of high temperature reactor excess heat in the process of coal and biomass co-gasification to hydrogen-rich gas

    International Nuclear Information System (INIS)

    The paper presents the results of the experimental study on the simulated application of HTR (High Temperature Reactor) excess heat in the process of allothermal co-gasification of coal and biomass. The laboratory scale installation with a fixed bed gasifier and auxiliary gasification agents pre-heating system, simulating the utilization of the HTR excess heat, were applied in the study. Steam and oxygen were the gasification media employed, and the process was focused on hydrogen-rich gas production. The results of the co-gasification of fuel blends of various biomass content at 800 °C and in various system configurations proved that the application of the simulated HTR excess heat in pre-heating of the gasification agents leads to the increase in the gaseous product yield. Furthermore, the HCA (Hierarchical Clustering Analysis) employed in the experimental data analysis revealed that the gasification of fuel blends of 20 and 40%w/w of biomass content results in higher volumes of the total gas, hydrogen, carbon monoxide and carbon dioxide than gasification of fuel blends of higher biomass content. - Highlights: • Simulated utilization of HTR excess heat in co-gasification of coal and biomass. • Assessment of three system configurations in terms of hydrogen production. • Application of the HCA in the experimental data set analysis. • Variation in gas components volume and content with fuel blend composition

  7. An experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    The relocation of degraded core material through the Control Rod Guide Tubes (CRGTs) is one of essential subjects to achieve the in-vessel retention (IVR) in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The CRGT is available as the discharge path by its failure in the core region and heat-transfer from the core-material to the CRGT is one of dominant factors in its failure. In case of a core design into which a fuel subassembly with an inner duct structure (FAIDUS) is introduced, a mixture of solid-fuel and liquid-steel is supposed to remain in the core region since the FAIDUS could effectively eliminate fuel in liquid-state from the core region. Therefore, the objective of the present study is to obtain experimental knowledge for the evaluation of heat-transfer from the mixture of solid-fuel and liquid-steel to the CRGT. In the present study, an experiment was conducted using Impulse Graphite Reactor which is an experimental facility in National Nuclear Center of the Republic of Kazakhstan. In the experiment, the mixture of solid-fuel and liquid-steel was generated by a low-power nuclear heating of fuel and transferring its heat to steel, and then, data to consider the heat-transfer characteristics from the mixture of solid-fuel and liquid-steel to the CRGT were obtained. The heat-transfer characteristic was revealed by evaluating thermocouple responses observed in the experiment. Through the present study, knowledge was obtained to evaluate heat-transfer from the remaining core-materials to the CRGT. (author)

  8. 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.

  9. Analysis of sodium/argon flow in anti-siphon equipment's curved pipeline of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    According to the study of CEFR anti-siphon equipment structure and its two-phase flow phenomenon, we can deeply understand its passive function which could be used to reduce the leakage of sodium. Compared to the experimental data,a numerical model can correctly describe this phenomenon of leakage processes. (authors)

  10. 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)

  11. Economic and environmental performance of oil transesterification in supercritical methanol at different reaction conditions: Experimental study with a batch reactor

    International Nuclear Information System (INIS)

    Highlights: • Influence of reaction parameters on FAME yields has been investigated. • The highest yield (93%) was achieved after 15 min at 350 °C and 12 MPa. • Models which predict with high certainty yields at different reaction conditions. • Economic and environmental performance of supercritical transesterification. • The lowest costs and impacts are always achieved at the highest yields. - Abstract: This study aims to investigate the influence of various reaction parameters (temperatures, working pressures and reaction time) on biodiesel yields and environmental and economic performance of rapeseed oil transesterification in supercritical methanol. Experiments were carried out in a laboratory-scale batch reactor. Results were statistically analysed and multiple regression models which describe and predict biodiesel yields with high certainty at different reaction conditions were provided. The highest biodiesel yield (93 wt%) was achieved at 350 °C and 12 MPa after 15 min of reaction. The lowest direct costs and life cycle environmental impacts (in terms of GHG emissions and fossil energy demand) are achieved at the highest yield due to the lowest oil consumption per unit of biodiesel produced. The results of sensitivity analysis showed that even at significantly lower oil feedstock prices this observation stands firm

  12. Development of experimental apparatus for evaluating corrosion resistance of cladding materials applied for advanced power reactor. 1

    Energy Technology Data Exchange (ETDEWEB)

    Inohara, Yasuto; Ioka, Ikuo; Fukaya, Kiyoshi; Tachibana, Katsumi; Suzuki, Tomio; Kiuchi, Kiyoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kuroda, Yuji; Miyamoto, Satoshi [Japan Atomic Power Co., Tokyo (Japan)

    2001-03-01

    On the development of cladding materials for advanced power reactors, it is important to clarify long performance and to control the compatibility to high temperature water at heat conducting surfaces under heavy irradiation. On the present study, the high temperature water loop with an autoclave was made for examining the corrosion behavior up to the super critical water range and for developing the simulation testing technique under irradiation in the hot cell. The loop is applicable to immersion tests in the temperature and pressure ranges up to 450degC and 25 MPa that are covered the surface temperature range of fuel claddings. One of the characteristics of this apparatus is a pair of sapphire windows of autoclave for in-situ observations, and a phase transition from water to super critical water conditions was clearly verified through these windows. In this apparatus, it is possible to control the temperature, pressure and Dissolved Oxygen (DO) within a fluctuations of few % on three phases, namely, water, steam and super critical water. (author)

  13. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  14. Tasks related to increase of RA reactor exploitation and experimental potential, Independent CO2 loop for cooling the samples irradiated in RA reactor vertical experimental channels, (I-IV), part I

    International Nuclear Information System (INIS)

    This volume contains the description of the design project of the head of the low-temperature coolant loops needed for cooling the samples to be irradiated in the RA vertical experimental channels. The thermal and mechanical calculations are included as well as calculation of antireactivity and activation of the construction materials. Cost estimation data are included as well. The drawings included are: head of the coolant loop; diagram of CO2 coolant temperature dependence; diagrams of weight of the loop tubes in the channels; axial distribution of the thermal neutron flux. Engineering drawings of two design solutions of the low-temperature loops with details are part of this volume

  15. Experimental evaluation of the production of the poisons Xe-135 and Sm-149 of the TRIGA Mark III reactor with mixed core, configuration No. 16 (Final report of the project); Evaluacion experimental de la produccion de los venenos Xe-135 y Sm-149 del reactor TRIGA Mark III con nucleo mixto, config. No. 16 (Informe final del proyecto)

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-11-15

    It was generated the concentration curve of the Xe{sup 135} (t) during the TRIGA Mark III reactor operation cycle, for a continuous irradiation of 72 h to 1 MW of thermal power, as well as the accumulation curve of the isotope after the shutdown, for the fuel configuration No. 16 in the thermal column. The maximum negative reactivities generated by the Xe{sup 135} for operation times greater than 60 h to 1 MW and after the reactor shutdown its were of 1.968 {+-} 0.15 dollars and 2.30 {+-} 0.15 dollars respectively. When comparing these results with those theoretically calculated we find differences of the order of 3.6% and 5.34% which are understood inside the experimental error that on the average was of 7.6%. The results before mentioned have an important application during the start up process of the Reactor, when analyzing the value of the weekly reactivity excess of the core and when is choice the pattern of bars to use for experiments of but of 2 h, where is required to minimize the temporary and space interferences of the neutron flux. (Author)

  16. Experimental investigations of local flow parameters near the walls of fuel elements of fast breeder reactors. Pt. 2

    International Nuclear Information System (INIS)

    The here presented material is the result of the first stage of experiments performed on the NEM-2 model, with the initial comparative values for evaluation of the effects of the geometry dimensions in the cluster upon the fluid-dynamic conditions being concerned in the main. The investigation of those effects (presence of displacement bodies, displacement of single rods or rod groups etc.) will be subject of further experimental works. (orig.)

  17. Experimental techniques and results related to high burn-up investigations at the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    Investigations of phenomena associated with extended or high burn-up comprise a considerable part of the fuel and materials research programme carried out at the OECD Halden Reactor Project, reflecting the interests and priorities of the nuclear industry in this area. For this purpose, sensors, re-instrumentation techniques, irradiation rigs and data evaluation methods have been developed and successfully applied which allow tests to be conducted at extended burn-up with high data quality. Re-instrumentation of base-irradiated fuel rods (50 MWd/kgU) with pressure transducers has been carried out with several rods, which were then subjected to high power, to study fission gas release. The results as well as those from another high burn-up test confirm the empirical fission gas release threshold developed at Halden, which predicts the onset of FGR to be about 1100 deg. C at this burn-up. The related interlinkage-resintering effects can be studied in special assemblies allowing sweep-out and analysis of fission products. Comprehensive investigations of fuel temperature data suggest a gradual degradation of UO2 thermal conductivity with burn-up. A special experiment is presently being irradiated to burn-up beyond 60 MWd/kgU, where the accuracy of temperature measurements is maintained by utilising expansion thermometers, insensitive to irradiation induced decalibration. A preliminary evaluation of the temperature increase points to a UO2 conductivity degradation of 35% at 50 MWd/kgUO2 and 600 deg. C. (author). 5 refs, 5 figs

  18. 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.

  19. 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.

  20. Experimental evaluation and modeling of liquid jet penetration to estimate droplet size in a three-phase riser reactor

    Institute of Scientific and Technical Information of China (English)

    Ali Akbar Jamali; Shahrokh Shahhosseini; Yaghoub Behjat

    2016-01-01

    In this work, the effects of injecting an evaporating liquid jet into solid–gas flow are experimentally investigated. A new model (SHED model) and a supplementary model (spray model) have also been proposed to investigate some flow-field characteristics in three-phase fluidized bed with the mean relative error 4.3%between model and measured results. Some experiments were conducted to study the influences of flow-field parameters such as liquid volumetric flow rate, injection velocity, jet angle and gas superficial velocity as well as solid mass flux on the jet penetration depth (JPD). In addition, independent variables were experimentally employed to propose two empirical correlations for JPD by using multiple regression method and spray cone angle (SCA) by using dimensional analysis technique. The mean relative errors between the JPD and SCA correlations versus ex-perimental data were 7.5%and 3.9%, respectively. In addition, in order to identify the variable effect, a parametric study was carried out. Applying the proposed model can avoid direct use of expensive devices to measure JPD and to predict droplet size.

  1. An experimental study of the liquid entrainment from swelled two-phase mixture surface in a reactor vessel

    International Nuclear Information System (INIS)

    An experimental study of liquid entrainment by rapid surface swellig of a two-phase mixture in a vessel has been performed. To investigate the effects of air flow and initial water level on the liquid entrainment, a series of experiments have been performed using air and water as working fluids. A total of 64 experimental liquid entrainment rate data have been obtained for various combinations of test parameters (i.e., six different initial water levels and varying air flow rates from 300 to 1,200 lpm) using two test vessels of the same height and different inner diameters (D=0.15 and 0.30m, respectively) for vertical bubbly and churn-turbulent flow conditions. An empirical correlation for the liquid entrainment rate, E, has been developed in terms of the superficial velocity of air, the initial water level, the density of gas, the surface tension, and the gravity. This correlation shows a good agreement with the present experimental data within 30% over a wide range of flow parameters

  2. Development and numerical/experimental characterization of a lab-scale flat flame reactor allowing the analysis of pulverized solid fuel devolatilization and oxidation at high heating rates

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, R., E-mail: romain.lemaire@mines-douai.fr; Menanteau, S. [Mines Douai, EI, F-59508 Douai (France)

    2016-01-15

    This paper deals with the thorough characterization of a new experimental test bench designed to study the devolatilization and oxidation of pulverized fuel particles in a wide range of operating conditions. This lab-scale facility is composed of a fuel feeding system, the functioning of which has been optimized by computational fluid dynamics. It allows delivering a constant and time-independent mass flow rate of fuel particles which are pneumatically transported to the central injector of a hybrid McKenna burner using a carrier gas stream that can be inert or oxidant depending on the targeted application. A premixed propane/air laminar flat flame stabilized on the porous part of the burner is used to generate the hot gases insuring the heating of the central coal/carrier-gas jet with a thermal gradient similar to those found in industrial combustors (>10{sup 5} K/s). In the present work, results issued from numerical simulations performed a priori to characterize the velocity and temperature fields in the reaction chamber have been analyzed and confronted with experimental measurements carried out by coupling particle image velocimetry, thermocouple and two-color pyrometry measurements so as to validate the order of magnitude of the heating rate delivered by such a new test bench. Finally, the main features of the flat flame reactor we developed have been discussed with respect to those of another laboratory-scale system designed to study coal devolatilization at a high heating rate.

  3. Development and numerical/experimental characterization of a lab-scale flat flame reactor allowing the analysis of pulverized solid fuel devolatilization and oxidation at high heating rates

    Science.gov (United States)

    Lemaire, R.; Menanteau, S.

    2016-01-01

    This paper deals with the thorough characterization of a new experimental test bench designed to study the devolatilization and oxidation of pulverized fuel particles in a wide range of operating conditions. This lab-scale facility is composed of a fuel feeding system, the functioning of which has been optimized by computational fluid dynamics. It allows delivering a constant and time-independent mass flow rate of fuel particles which are pneumatically transported to the central injector of a hybrid McKenna burner using a carrier gas stream that can be inert or oxidant depending on the targeted application. A premixed propane/air laminar flat flame stabilized on the porous part of the burner is used to generate the hot gases insuring the heating of the central coal/carrier-gas jet with a thermal gradient similar to those found in industrial combustors (>105 K/s). In the present work, results issued from numerical simulations performed a priori to characterize the velocity and temperature fields in the reaction chamber have been analyzed and confronted with experimental measurements carried out by coupling particle image velocimetry, thermocouple and two-color pyrometry measurements so as to validate the order of magnitude of the heating rate delivered by such a new test bench. Finally, the main features of the flat flame reactor we developed have been discussed with respect to those of another laboratory-scale system designed to study coal devolatilization at a high heating rate.

  4. 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).

  5. Development and numerical/experimental characterization of a lab-scale flat flame reactor allowing the analysis of pulverized solid fuel devolatilization and oxidation at high heating rates.

    Science.gov (United States)

    Lemaire, R; Menanteau, S

    2016-01-01

    This paper deals with the thorough characterization of a new experimental test bench designed to study the devolatilization and oxidation of pulverized fuel particles in a wide range of operating conditions. This lab-scale facility is composed of a fuel feeding system, the functioning of which has been optimized by computational fluid dynamics. It allows delivering a constant and time-independent mass flow rate of fuel particles which are pneumatically transported to the central injector of a hybrid McKenna burner using a carrier gas stream that can be inert or oxidant depending on the targeted application. A premixed propane/air laminar flat flame stabilized on the porous part of the burner is used to generate the hot gases insuring the heating of the central coal/carrier-gas jet with a thermal gradient similar to those found in industrial combustors (>10(5) K/s). In the present work, results issued from numerical simulations performed a priori to characterize the velocity and temperature fields in the reaction chamber have been analyzed and confronted with experimental measurements carried out by coupling particle image velocimetry, thermocouple and two-color pyrometry measurements so as to validate the order of magnitude of the heating rate delivered by such a new test bench. Finally, the main features of the flat flame reactor we developed have been discussed with respect to those of another laboratory-scale system designed to study coal devolatilization at a high heating rate. PMID:26827350

  6. Water reactor fuel element modelling at high burnup and its experimental support. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The Technical Committee Meeting on Fuel Element Modelling at High Burnup and its Experimental Support was recommended by the International Working Group on Fuel Performance and Technology (IWGFPT). Its subject had been touched on in many of the IAEA's activities; however for the first time modellers and experimentalists were brought together to have an exchange of views on the research under way and to identify areas where new knowledge is necessary to improve the safety, reliability and/or economics of nuclear fuel. The timely organization of this meeting in conjunction with the second meeting of the Co-ordinated Research Programme on Fuel Modelling at Extended Burnup, in short ''FUMEX'', allowed fruitful participation of representatives of developing countries which are only rarely exposed to such a scientific event. The thirty-nine papers presented covered the status of codes and experimental facilities and the main phenomena affecting the fuel during irradiation, namely: thermal fuel performance, clad corrosion and pellet-cladding interaction (PCI) and fission gas release (FGR). Refs, figs, tabs

  7. Experimental and modeling study of the oxidation of n-butane in a jet stirred reactor using cw-CRDS measurements.

    Science.gov (United States)

    Bahrini, Chiheb; Morajkar, Pranay; Schoemaecker, Coralie; Frottier, Ophélie; Herbinet, Olivier; Glaude, Pierre-Alexandre; Battin-Leclerc, Frédérique; Fittschen, Christa

    2013-12-01

    The gas-phase oxidation of n-butane has been studied in an atmospheric jet-stirred reactor (JSR) at temperatures up to 950 K. For the first time, continuous wave cavity ring-down spectroscopy (cw-CRDS) in the near-infrared has been used, together with gas chromatography (GC), to analyze the products formed during its oxidation. In addition to the quantification of formaldehyde and water, which is always difficult by GC, cw-CRDS allowed as well the quantification of hydrogen peroxide (H2O2). A comparison of the obtained mole fraction temperature profiles with simulations using a detailed gas-phase mechanism shows a good agreement at temperatures below 750 K, but an overestimation of the overall reactivity above this temperature. Also, a strong overestimation was found for the H2O2 mole fraction at higher temperatures. In order to improve the agreement between model and experimental results, two modifications have been implemented to the model: (a) the rate constant for the decomposition of H2O2 (+M) ↔ 2OH (+M) has been updated to the value recently proposed by Troe (Combust. Flame, 2011, 158, 594-601) and (b) a temperature dependent heterogeneous destruction of H2O2 on the hot reactor walls with assumed rate parameters has been added. The improvement (a) slows down the overall reactivity at higher temperatures, but has a negligible impact on the maximal H2O2 mole fraction. Improvement (b) has also a small impact on the overall reactivity at higher temperatures, but a large effect on the maximal H2O2 mole fraction. Both modifications lead to an improved agreement between model and experiment for the oxidation of n-butane in a JSR at temperatures above 750 K. PMID:24135810

  8. Experimental investigation of the behaviour of a fibre-type thermal insulation for high temperature reactors under rapid depressurization

    International Nuclear Information System (INIS)

    Thermal insulations in high temperature nuclear power plants must withstand very rapid depressurization rates, that occur during normal operation following a removal of the generator load or in an emergency situation such as a recuperator or turbomachinery failure. Rapid depressurization causes transient pressure differences across the hot internal structural elements of the insulation, that consequently can lead to failure. The authors report the results of an experimental study of the depressurization on a particular model insulation. The rate of depressurization has been varied from 12 bar/s to 4000 bar/s. The maximum pressure differences over the internal structures as a function of the depressurization rate show characteristic behaviour. The pressure differences remain very small up to a certain characteristic depressurization rate and increase then rapidly with increasing rates. (Auth.)

  9. 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.

  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

    International Nuclear Information System (INIS)

    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. 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)

  12. Conceptual design of in-vessel viewing and inspection system (IVIS) for international thermonuclear experimental reactor (ITER)

    International Nuclear Information System (INIS)

    Visual information on in-vessel components of ITER, such as divertor and armor tile, is taken by in-vessel viewing and inspection system (IVIS). The IVIS is consist of a visual sensor, a manipulator, a control system and a data processing system. In the IVIS, the sensor and the manipulator are operated under high temperature (about 150degC), high vacuum (about 10-8 Torr) and high radiation (about 107 R/hr) conditions. In this report, we studied three different concepts of the IVIS using a fiberscope, a periscope and a industrial television (ITV), respectively. The ITV concept is simple and has high mobility, but radiation, heat and magnetic resistances of ITV are critical issues to apply it to the fusion device, such as ITER. On the other hand, the periscope concept is able to get a large visual field, and the radiation resistance is better than those another concepts. But, this concept has some dead angles by geometrical limitations of the mechanism. Feature of the fiberscope concept is intermediate between those of ITV and periscope ones. From conceptual design of the IVIS, we decided to carry out the irradiation tests of critical elements of fiberscope and periscope. The experimental results become the data base the engineering design of IVIS. (author)

  13. N Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The last of Hanfordqaodmasdkwaspemas7ajkqlsmdqpakldnzsdflss nine plutonium production reactors to be built was the N Reactor.This reactor was called a dual purpose...

  14. 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.)

  15. 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)

  16. Household energy consumption: the future is in our hands. ITER, an experimental fusion reactor. Do CO2-free energies exist? Liquefied natural gas, king of the gas market

    International Nuclear Information System (INIS)

    This issue of Alternatives newsletter features 4 main articles dealing with: 1 - Household energy consumption - the future is in our hands: With energy resources growing scarcer and more expensive, everyone has a duty to conserve energy. Because combating global warming also means adopting simple habits and using the right equipment - with help from our governments to lead us to change. A practical look at what we can do. 2 - ITER, an experimental fusion reactor: The entire international community is trying to reproduce here on Earth the fusion of hydrogen atoms occurring naturally in the Sun, lured by the promise of a virtually inexhaustible source of energy. More on ITER from the project's Director General. 3 - Do CO2-free energies exist?: As nations struggle to reduce greenhouse gas emissions, the question is moot. Environmental engineer Jean-Marc Jancovici gives us his point of view. 4 - Liquefied natural gas, king of the gas market: LNG's many advantages are enticing industry to develop supply routes and infrastructure to meet strong demand. But the race for LNG is not without its limits

  17. Experimental and theoretical study of a particular type of transient flow of boiling sodium: flow excursion. Study carried out in the framework of the fast neutron reactor safety

    International Nuclear Information System (INIS)

    The flow excursion phenomenon was studied in the framework of safety analyses for sodium cooled fast reactors, as a consequence of an accident of pump breakdown without safety rod drop. The experimental study, performed on an out-of-pile circuit with sodium forced convection allowed the flow excursion due to boiling appearance in a single heating channel feeded in parallel with a by-pass to be investigated. The first phase is characterized by a relatively slow decay of the (mean) flow rate through the channel from its initial value (corresponding to starting boiling) up to a mean value near zero; its duration, rather important, is of the order of ten or thirty seconds. The second phase appears when the boiling zone occupies an important fraction (half or three quarters) of the heating length of the channel. The mean flow rate has a low value, near zero; the flow rate is oscillatory (chugging) and is formed of vapor plugs parted by liquid bundles. This chugging phase lasts two or three seconds. The computation model developed described simply the boiling liquid flow (homogeneous sliding model) and the heat transfer between the heating medium and the fluid. The transient terms are neglected in continuity and energy equations so as the fluid in the channel is incompressible and its thermal inertia nul

  18. Methodological developments and qualification of calculation schemes for the modelling of photonic heating in the experimental devices of the future Jules Horowitz material testing reactor (RJH)

    International Nuclear Information System (INIS)

    The objective of this work is to develop the modelling of the nuclear heating of the experimental devices of the future Jules Horowitz material testing reactor (RJH). The strong specific nuclear power produced (460 kW/l), induces so intense photonic fluxes which cause heating and large temperature gradients that it is necessary to control it by an adequate design. However, calculations of heating are penalized by the very large uncertainties estimated at a value of about 30% (2*σ) coming from the gaps and uncertainties of the data of gamma emission present in the libraries of basic nuclear data. The experimental program ADAPh aims at reducing these uncertainties. Measurements by thermoluminescent detectors (TLD) and ionisation chambers are carried out in the critical assemblies EOLE (Mox) and Minerve (UO2). The rigorous interpretation of these measurements requires specific developments based on Monte-Carlo simulations of coupled neutron-gamma and gamma-electron transport. The developments carried out are made different in particular by the modelling of cavities phenomena and delayed gamma emissions by the decay of fission products. The comparisons calculation-measurement made it possible to identify a systematic bias confirming a tendency of calculations to underestimate measurements. A Bayesian method of adjustment was developed in order to re-estimate the principal components of the gamma heating and to transpose the results obtained to the devices of the RJH, under conditions clearly and definitely representative. This work made possible to reduce significantly the uncertainties on the determination of the gamma heating from 30 to 15 per cent. (author)

  19. Computational Fluid Dynamics (CFD) for Nuclear Reactor Safety Applications - Workshop Proceedings, CFD4NRS-3 - Experimental Validation and Application of CFD and CMFD Codes to Nuclear Reactor Safety Issues

    International Nuclear Information System (INIS)

    The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to WGAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. The workshop included single-phase and multiphase CFD applications as well as new experimental techniques, including the following: Single-phase and two-phase CFD simulations with an emphasis on validation were sought in areas such as boiling flows, free-surface flows, direct contact condensation, and turbulent mixing. These should relate to NRS-relevant issues such as pressurized thermal shock, critical heat flux, pool heat exchangers, boron dilution, hydrogen distribution, and thermal striping. The use of systematic error quantification and Best Practice Guidelines (BPGs) was encouraged. Experiments providing data suitable for CFD validation-specifically in the area of NRS-including local measurement devices such as multi-sensor optical or electrical probes, Laser Doppler Velocimetry (LDV), hot-film/wire anemometry, Particle Image Velocimetry (PIV), Laser-Induced Fluorescence (LIF), and other innovative techniques. There were over 200 registered participants at the CFD4NRS-3 workshop. The program consisted of about 75 technical papers. Of these, 57 were oral presentations and 19 were posters. An additional 20 posters related to the OECD/NEA-sponsored CFD benchmark exercise on thermal fatigue in a T-Junction were presented. In addition, five keynote lectures were given by distinguished experts. This is about a 30 pc increase with respect to the previous XCFD4NRS workshop held in Grenoble in 2008, and a 70 pc increase compared to the first CFD4NRS workshop held in Garching in 2006. This confirms that there is a real and growing need for such workshops. The papers presented in the conference tackled different topics

  20. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 1014 n/cm2/sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  1. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Science.gov (United States)

    Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

    2015-08-01

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  2. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  3. 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.

  4. For a Better Estimation of Gamma Heating in Experimental Reactors and Devices: Stakes and Work Plan from Calculation Methods to Nuclear Data

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is an international Material-Testing Reactor currently under construction at CEA Cadarache. The determination of gamma heating levels in this future commercial reactor is of crucial importance as gamma heating affects both safety and performance parameters of the JHR. Required accuracy (5% at one standard deviation) makes it necessary to calibrate bias and uncertainty associated with JHR gamma-heating calculations. Main steps of bias determination for gamma-heating calculations include, firstly, the development of a calculation methodology with controlled use of physical approximations; secondly, the interpretation of gamma-heating measurements so as to determine bias supposed to be only due to nuclear data. (authors)

  5. Test reactor technology

    International Nuclear Information System (INIS)

    The Reactor Development Program created a need for engineering testing of fuels and materials. The Engineering Test Reactors were developed around the world in response to this demand. The design of the test reactors proved to be different from that of power reactors, carrying the fuel elements closer to the threshold of failure, requiring more responsive instrumentation, more rapid control element action, and inherent self-limiting behavior under accident conditions. The design of the experimental facilities to exploit these reactors evolved a new, specialized, branch of engineering, requiring a very high-lvel scientific and engineering team, established a meticulous concern with reliability, the provision for recovery from their own failures, and detailed attention to possible interactions with the test reactors. This paper presents this technology commencing with the Materials Testing Reactor (MTR) through the Fast Flux Test Facility, some of the unique experimental facilities developed to exploit them, but discusses only cursorily the experiments performed, since sample preparation and sample analyses were, and to some extent still are, either classified or proprietary. The Nuclear Engineering literature is filled with this information

  6. 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.

  7. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

    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

  8. Analysis of higher power research reactors' parameters

    International Nuclear Information System (INIS)

    The objective of this monograph was to analyze and compare parameters of different types of research reactors having higher power. This analysis could be used for decision making and choice of a reactor which could possibly replace the existing ageing RA reactor in Vinca. Present experimental and irradiation needs are taken into account together with the existing reactors operated in our country, RB and TRIGA reactor

  9. RB research nuclear reactor - Annual report for 1986, I - III

    International Nuclear Information System (INIS)

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986

  10. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

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

  11. Optimal control of nuclear reactors

    International Nuclear Information System (INIS)

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

  12. Experimental study on the operating characteristics of an inner preheating transpiring wall reactor for supercritical water oxidation: Temperature profiles and product properties

    International Nuclear Information System (INIS)

    A new process to generate multiple thermal fluids by supercritical water oxidation (SCWO) was proposed to enhance oil recovery. An inner preheating transpiring wall reactor for SCWO was designed and tested to avoid plugging in the preheating section. Hot water (400–600 °C) was used as auxiliary heat source to preheat the feed to the reaction temperature. The effect of different operating parameters on the performance of the inner preheating transpiring wall reactor was investigated, and the optimized operating parameters were determined based on temperature profiles and product properties. The reaction temperature is close to 900 °C at an auxiliary heat source flow of 2.79 kg/h, and the auxiliary heat source flow is determined at 6–14 kg/h to avoid the overheating of the reactor. The useful reaction time is used to quantitatively describe the feed degradation efficiency. The outlet concentration of total organic carbon (TOCout) and CO in the effluent gradually decreases with increasing useful reaction time. The useful reaction time needed for complete oxidation of the feed is 10.5 s for the reactor. - Highlights: • A new process to generate multiple thermal fluids by SCWO was proposed. • An inner preheating transpiring wall reactor for SCWO was designed and tested. • Hot water was used as auxiliary heat source to preheat the feed at room temperature. • Effect of operating parameters on the performance of the reactor was investigated. • The useful reaction time required for complete oxidation of the feed is 10.5 s

  13. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    1990-06-01

    The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

  14. Safety of research reactors

    International Nuclear Information System (INIS)

    The number of research reactors that have been constructed worldwide for civilian applications is about 651. Of the reactors constructed, 284 are currently in operation, 258 are shut down and 109 have been decommissioned. More than half of all operating research reactors worldwide are over thirty years old. During this long period of time national priorities have changed. Facility ageing, if not properly managed, has a natural degrading effect. Many research reactors face concerns with the obsolescence of equipment, lack of experimental programmes, lack of funding for operation and maintenance and loss of expertise through ageing and retirement of the staff. Other reactors of the same vintage maintain effective ageing management programmes, conduct active research programmes, develop and retain high calibre personnel and make important contributions to society. Many countries that operate research reactors neither operate nor plan to operate power reactors. In most of these countries there is a tendency not to create a formal regulatory body. A safety committee, not always independent of the operating organization, may be responsible for regulatory oversight. Even in countries with nuclear power plants, a regulatory regime differing from the one used for the power plants may exist. Concern is therefore focused on one tail of a continuous spectrum of operational performance. The IAEA has been sending missions to review the safety of research reactors in Member States since 1972. Some of the reviews have been conducted pursuant to the IAEA' functions and responsibilities regarding research reactors that are operated within the framework of Project and Supply Agreements between Member States and the IAEA. Other reviews have been conducted upon request. All these reviews are conducted following procedures for Integrated Safety Assessment of Research Reactors (INSARR) missions. The prime objective of these missions has been to conduct a comprehensive operational safety

  15. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    Science.gov (United States)

    Zweibaum, Nicolas

    The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). (Abstract shortened by UMI.).

  16. Comparison of activation cross section measurements and experimental techniques for fusion reactor technology. Summary report of the IAEA specialists' meeting held at the Tokai Research Establishment, JAERI, Japan, 15 to 17 November 1993

    International Nuclear Information System (INIS)

    The present report contains the Summary of the IAEA Specialists' Meeting (SPM) on ''Comparison of Activation Cross Section Measurements and Experimental Techniques for Fusion Reactor Technology'', held at the Tokai Research Establishment, JAERI, Japan, from 15 to 17 November 1993. This SPM was organized by the IAEA Nuclear Data Section (NDS) with the co-operation and assistance of the Japan Atomic Energy Research Institute. The purpose of the Specialists' Meeting was to form an international programme of the comparison of activation cross section measurements and experimental techniques useful in reactor technology. It was agreed that new activation measurements be performed at the Fusion Neutronics Source Facility (FNS) of JAERI, neutron source facilities of the V.G. Khlopin Radium Institute, St. Petersburg, Russia, and in the Institute of Experimental Physics, Debrecen, Hungary. A list of the most important reactions has been prepared. Fourteen experts from seven Member States attended the meeting. The conclusions and recommendations are presented on the basis of discussions held by meeting participants. (author). 4 tabs

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

    International Nuclear Information System (INIS)

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

  18. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors; Etude experimentale des instabilites hydrodynamiques survenant dans les reacteurs nucleaires a ebullition

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-10-01

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [French] II a ete effectue une etude experimentale, en boucle hors-pile, des oscillations hydrodynamiques survenant dans les reacteurs a ebullition. L'etude a ete effectuee a la pression atmospherique et a une pression voisine de 8 atmospheres dans des canaux chauffes electriquement a puissance imposee constante et uniforme. Dans les essais a 8 atmospheres le canal etait un tube circulaire de diametre interieur 6 mm environ. A 1 atmosphere le canal etait de section annulaire 10 * 20 mm avec un tube interieur chauffant et un tube exterieur en pyrex. Le fonctionnement etait possible

  19. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

  20. Fusion reactor materials

    International Nuclear Information System (INIS)

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

  1. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

    This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

  2. Comparison between experimental and computational measures of the sample GU3 irradiated in The Goscen NPP reactor; Comparacion entre las medidas experimentales y computacionales de la muestra GU3 irradiada en el reactor Goscen NPP

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Rivada, A.; Tore, C.

    2014-07-01

    For improvement and validation of neutron codes with calculated burning requires experimental measurements of the isotope inventory of irradiated fuel. The quality of these codes of calculation is fundamental to the security of the transport and storage of the same studies. The comparison was conducted between the calculated values and experimental measures MONTEBURNS and SCALE of the sample codes GU3. (Author)

  3. Reactor building

    International Nuclear Information System (INIS)

    The whole reactor building is accommodated in a shaft and is sealed level with the earth's surface by a building ceiling, which provides protection against penetration due to external effects. The building ceiling is supported on walls of the reactor building, which line the shaft and transfer the vertical components of forces to the foundations. The thickness of the walls is designed to withstand horizontal pressure waves in the floor. The building ceiling has an opening above the reactor, which must be closed by cover plates. Operating equipment for the reactor can be situated above the building ceiling. (orig./HP)

  4. Fast Reactor Development Strategy in China

    International Nuclear Information System (INIS)

    As one of the largest developing countries, China needs a reliable energy supplement. At the same time, China should improve the energy structure to decrease CO2 emissions. Nuclear and renewable energies are the main solutions to these issues. According to the research results, the nuclear capacity should increase to 400 GW(e) up to 2050. Fast reactors must be developed considering the limitation of uranium resources. In order to deploy fast reactor technology, the ‘experimental reactor, demonstration reactor and commercial reactor’ strategy has been suggested. China has finished the construction of the China Experimental Fast Reactor (CEFR) and gained necessary experience about fast reactors. The China Institute of Atomic Energy (CIAE) has begun to design the CFR-600, a 600 MW(e) demonstration fast reactor. This reactor will be put into operation before 2025. After that, a larger commercial reactor will be constructed. Besides fast reactors, all of other key sectors of fuel cycle will be developed at the same time such as reprocessing, fast reactor fuel, etc. There are two main tasks of fast reactors, one of which is to raise the utility ratio of uranium, and the other one is to transmute the long life waste of light water reactors. The fast reactor will be designed as a breeder and burner, respectively. (author)

  5. 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)

  6. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-06-26

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  7. Experimental Plan and Irradiation Target Design for FeCrAl Embrittlement Screening Tests Conducted Using the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    The objective of the FeCrAl embrittlement screening tests being conducted through the use of Oak Ridge National Laboratories (ORNL) High Flux Isotope Reactor is to provide data on the radiation-induced changes in the mechanical properties including radiation-induced hardening and embrittlement through systematic testing and analysis. Data developed on the mechanical properties will be supported by extensive microstructural evaluations to assist in the development of structure-property relationships and provide a sound, fundamental understanding of the performance of FeCrAl alloys in intense neutron radiation fields. Data and analysis developed as part of this effort will be used to assist in the determination of FeCrAl alloys as a viable material for commercial light water reactor (LWR) applications with a primary focus as an accident tolerant cladding.

  8. Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

    Energy Technology Data Exchange (ETDEWEB)

    Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.; Nasonov, V. A.; Vihrov, V. I.; Erak, D. Yu. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

  9. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  10. Automated reactor records evaluation framework

    International Nuclear Information System (INIS)

    The only truly reliable method for core physics code validation is comparison against experimental data – and for power nuclear reactors, the only reasonably acquirable kind of experimental data are the reactor records. However, the amount of the data coming from the reactor operation is often so vast that it can be discouraging for the code developers to use it properly. Thus, the validation package is further reduced because the data is hard to use. This paper presents an elaborate, fully automated framework, which was designed and implemented in our institute, for reactor records processing and its use for core physics code validation. The workflow, implemented as a Web 2.0 application, provides a practical and painless solution for use of reactor records data for code development and validation. (author)

  11. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  12. 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.

  13. 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)

  14. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  15. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

    Current work is described for Compton Recoil Gamma-Ray Spectrometry including developments in experimental technique as well as recent reactor spectrometry measurements. The current status of the method is described concerning gamma spectromoetry probe design and response characteristics. Emphasis is given to gamma spectrometry work in US LWR and BR programs. Gamma spectrometry in BR environments are outlined by focussing on start-up plans for the Fast Test Reactor (FTR). Gamma spectrometry results are presented for a LWR pressure vessel mockup in the Poolside Critical Assembly (PCA) at Oak Ridge National Laboratory

  16. 2 MW液态钍基熔盐实验堆主屏蔽温度场分析%Temperature field analysis for the main shielding of the 2-MW thorium-based molten salt experimental reactor

    Institute of Scientific and Technical Information of China (English)

    何杰; 夏晓彬; 蔡军; 潘登; 彭玉; 黄建平; 张国庆

    2016-01-01

    Background:Molten salt reactor is a fourth generation advanced reactor. The concrete wall is the key part of this high-temperature reactor shielding, so temperature field analysis is important. Purpose: This study attempts to calculate the temperature field of the TMSR-LF1 (2-MW liquid-fueled molten salt experimental reactor) shielding, and judge if it meets the design requirements. Methods: In accordance with the problem that MCNP (Monte Carlo N Particle Transport Code) results cannot be directly imported into Fluent, a program which converts MCNP results to the spatial distribution of power density, and imports the spatial distribution of power density into the Fluent in the form of User-Defined Function (UDF) was developed by using Python programming language to realize the coupling of the two. According to TMSR-LF1 design parameters, a one-eight physical and thermal model of the whole reactor is established, using code MCNP and Fluent. Reactor radiation shielding thermal analysis adopts the assumptions that the different environment temperatures are 5°C, 18°C, 25°C, 30°C, 35°C and 40°C, respectively.Results: The maximal values of temperature and temperature gradient in the radiation shielding concrete wall are 67.42 °C and 78.40 °C·m?1, which are lower than limit values.Conclusion: The radiation shielding concrete wall can meet the design requirements.%反应堆主屏蔽是核反应堆的重要组成部分,用来有效降低反应堆运行时屏蔽体外的辐射剂量水平,以满足反应堆部件材料对辐射限制的要求.温度是影响反应堆主屏蔽性能的重要因素.针对2 MWth液态熔盐堆(2-MW liquid-fueled molten salt experimental reactor,TMSR-LF1),采用MCNP软件获得功率分布后,利用Fluent软件对主屏蔽进行温度场计算.计算过程中利用Python语言编写了程序(MCNP to Fluent,MTF)来实现将MCNP(Monte Carlo N Particle Transport Code)计算结果转换为功率密度的空间分布,

  17. The Glucose Water Distribution Start UASB Reactor Experimental Study%葡萄糖配水启动UASB反应器实验研究

    Institute of Scientific and Technical Information of China (English)

    梁贤军; 张欣; 李尚科; 白晶

    2013-01-01

    the pilot study is effective in a single volume of 31L UASB reactor, with pig manure as culture, reactor tem-perature to room temperature, the water for glucose water, control water pH value in 6.5~7.5 between, after 135 days of launch, CODCr volume load from the 0.5kg/ ( m3.d ) to 4.0KG ( m3.d ), CODCr removal rate can reach over 90%. Be-cause the start time is not long, the reactor has appeared only sludge flocs, granular sludge has not appeared yet, start al-most completed.%  该试验研究是在一个有效容积为31L的UASB反应器中进行的,以猪粪为菌种,反应器温度为常温,试验用水为葡萄糖配水,控制进水pH值在6.5~7.5之间,经过135 d的启动,CODCr的容积负荷由最初的0.5 kg/(m3.d)增至4.0 kg/(m3.d),CODCr去除率达到90%以上。由于启动时间不长,反应器内只出现了污泥絮体,颗粒污泥尚未出现,启动基本完成。

  18. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    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

  19. Experimental study of void behavior in a suppression pool of a boiling water reactor during the blowdown period of a loss of coolant accident

    Science.gov (United States)

    Rassame, Somboon

    The possible failure of an Emergency Core Cooling System (ECCS) train due to a large amount of entrained gas in the ECCS pump suction piping in a Loss of Coolant Accident (LOCA) is one of the potential engineering problems faced in a Boiling Water Reactor (BWR) power plant. To analyze potential gas intrusion into the ECCS pump suction piping, the study of void behavior in the Suppression Pool (SP) during the LOCA is necessary. The void fraction distribution and void penetration are considered as the key parameters in the problem analysis. Two sets of experiments, namely, steady-state tests and transient tests were conducted using the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) to study void behavior in the SP during the blowdown. The design of the test apparatus used is based on the scaling analysis from a prototypical BWR containment (MARK-I) with consideration of the downcomer size, the SP water level, and the downcomer water submergence depth. Several instruments were installed to obtain the required experimental data, such as inlet gas volumetric flow, void fraction, pressure, and temperature. For the steady-state tests, the air was injected through a downcomer pipe in the SP in order to simulate the physical phenomena in the SP during the initial blowdown of LOCA. Thirty tests were performed with two different downcomer sizes (0.076 and 0.102 m), various air volumetric flow rates or flux (0.003 to 0.153 m3/s or 0.5 to 24.7 m/s), initial downcomer void conditions (fully filled with water, partially void, and completely void) and air velocity ramp rates (one to two seconds). Two phases of the experiment were observed, namely, the initial phase and the quasi-steady phase. The initial phase produced the maximum void penetration depth; and the quasi-steady phase showed less void penetration with oscillation in the void penetration. The air volumetric flow rate was found to have a minor effect on the void fraction

  20. 白腐真菌生物膜反应器中活性艳红 X-3B脱色与降解的实验研究%Experimental Study on Decolorization and Degradation of Reactive Brilliant Red X-3B in a White Rot Fungal Biofilm Reactor

    Institute of Scientific and Technical Information of China (English)

    黄民生; 黄荣; 程永前; 张国莹

    2001-01-01

    Experimental results of an azo dye(reactive brilliant red X-3B, RBR X-3B) decolorization and degradation in a white rot fungal biofilm reactor were introduced and discussed. The fungal biofilm reactor is highly potential for dye decolorization and degradation with the highest decoloring rate of 95% within 96 hours reaction time at initial pH 4.5 under high nitrogen level (HN) (24 mmol/L ammonium tartrate) condition. Experimental conditions, such as nutrient nitrogen levels in reaction mixture and initial pH, significantly affected dye decolorization and degradation. Effluents from this biofilm reactor can be well treated to meet the discharging requirements by use of chemical flocculation. RBR X-3B was first absorbed onto fungal biomass and then degraded gradually. The SH-13 fungus monopolized the biofilm throughout the experiments, though the reactor was exposed to open air for 4 months.

  1. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  2. Experimental determination of spectral ratios and of neutrons energy spectrum in the fuel of the IPEN/MB-01 nuclear reactor

    International Nuclear Information System (INIS)

    This study aims to determine the spectral ratios and the neutron energy spectrum inside the fuel of IPEN/MB-01 Nuclear Reactor. These parameters are of great importance to accurately determine spectral physical parameters of nuclear reactors like reaction rates, fuel lifetime and also security parameters such as reactivity. For the experiment, activation detectors in the form of thin metal foils were introduced in a collapsible fuel rod. Then the rod was placed in the central position of the core which has a standard rectangular configuration of 26 x 28 fuel rods. There were used activation detectors from different elements such Au-197, U-238, Sc-45, Ni-58, Mg-24, Ti-47 and In-115 to cover a large range of the neutron energy spectrum. After the irradiation, the activation detectors were submitted to gamma spectrometry using a counting system with high purity Germanium, to obtain the reaction rates (saturation activity) per target nucleus. The spectral ratios were compared with calculated values obtained by the Monte Carlo method using the MCNP-4C code. The neutron energy spectrum was obtained inside the fuel rod using the SANDBP code with an input spectrum obtained by the MCNP-4C code, based on the saturation activity per target nucleus values of the activation detectors irradiated. (author)

  3. Operating experiences of the research reactors

    International Nuclear Information System (INIS)

    Nuclear research reactors are devices of wide importance, being used for different scientific research tasks, for testing and improving reactor systems and components, for the production of radioisotopes, for the purposes of defence, for staff training and for other purposes. There are three research reactors in Yugoslavia: RA, RB and TRIGA. Reactors RA and RB at the 'Boris Kidric' Institute of Nuclear Sciences are of heavy water type power being 6500 and 10 kW, and maximum thermal neutron flux of 1014 and 1011(n/cm2s), respectively. TRIGA reactor at the 'Jozef Stefan' Institute in Ljubljana is of 250 kW power and maximum thermal neutron flux of 1013(n/cm2s). Reactors RA and RB use soviet fuel in the form of uranium dioxide (80% enriched) and metallic uranium (2%). Besides, RB reactor operates with natural uranium too. TRIGA reactor uses american uranium fuel 70% and 20% enriched, uranium being mixed homogeneously with moderator (ZrH). Experiences in handling and controlling the fuel before irradiation in the reactor, in reactor and after it are numerous and valuable, involving either the commercial arrangements with foreign producers, or optimal burn up in reactor or fuel treatment after the reactor irradiation. Twenty years of operating experience of these reactors have great importance especially having in mind the number of trained staff. Maintenance of reactors systems and fluids in continuous operation is valuable experience from the point of view of water reactor utilization. The case of the RA reactor primary cycle cobalt decontamination and other events connected with nuclear and radiation security for all three reactors are also specially emphasized. Owing to our research reactors, numerous theoretical, numerical and experimental methods are developed for nuclear and other analyses and design of research and power reactors,as well as methods for control and protection of radiation. (author)

  4. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  5. 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.

  6. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  7. 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

  8. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  10. Final Safety Analysis Addenda to Hazards Summary Report, Experimental Breeder Reactor II (EBR-II): upgrading of plant protection system. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    Allen, N. L.; Keeton, J. M.; Sackett, J. I. [comps.

    1980-06-01

    This report is the second in a series of compilations of the formal Final Safety Analysis Addenda (FSAA`s) to the EBR-II Hazard Summary Report and Addendum. Sections 2 and 3 are edited versions of the original FSAA`s prepared in support of certain modifications to the reactor-shutdown-system portion of the EBR-II plant-protection system. Section 4 is an edited version of the original FSAA prepared in support of certain modifications to a system classified as an engineered safety feature. These sections describe the pre- and postmodification system, the rationale for the modification, and required supporting safety analysis. Section 5 provides an updated description and analysis of the EBR-II emergency power system. Section 6 summarizes all significant modifications to the EBR-II plant-protection system to date.

  11. Experimental and numerical study of long-term cooling of VVER-640 reactor in the Pactel facility using thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Banati, J.; Virtanen, E.; Purhonen, H. [Lappeenranta University of Technology, Lappeenranta (Finland); Alexandrine, S. [Scientific and Engineering Center for Nucl. and Radiation Safety, Moscow (Russian Federation); Volkova, S.N. [Alexandrov Research Institute of Technology, NITI, Sosnovy Bor (Russian Federation)

    2001-07-01

    The present paper is a result of joint efforts of Finnish and Russian research institutes in evaluation of the passive safety features of the new VVER-640 reactor design. For better understanding of the complex thermal hydraulic phenomena during the long-term cooling in the final stage of a large break LOCA, experiments were carried out in the PACTEL facility. In order to increase the confidence with numerical methods and to contribute to the validation process, the transients were simulated by computer codes, such as the RELAP, APROS, CATHARE and KORSAR. The calculated results are compared to the measured database and conclusions are drawn for the modeling abilities of the system codes to reproduce the main features of the transients. The current study is extended with discussion of various modeling aspects, characterization of the natural circulation, as well as code-specific nodalization problems. (authors)

  12. 中国实验快堆主蒸汽系统优化设计及分析研究%Optimization design and analysis on main steam system of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    纪西胜; 吴强; 牛敬娟

    2012-01-01

    The function of the main steam system of China Experimental Fast Reactor (CEFR) is to transfer the steam from SG to the turbine to generate power and to discharge the steam from reactor in case of accidental condition. However, the reactor automatically shut down many times due to improper operation of valves, affecting the stability of the system and increasing operation cost. In order to optimize the steam system process flow, this paper introduced the bypass and worked out the design parameters, and finally gave qualitative analysis of the calculation result in special condition.%中国实验快堆三回路主蒸汽系统主要功能是将蒸汽发生器产生的蒸汽送至汽轮发电机组,辅助功能是在事故工况下排出反应堆产生的热量.调试期间多次因主蒸汽系统阀门手动操作而引起停堆,影响了系统的稳定性,增加了运行成本.本文对主蒸汽系统进行了优化设计、增加旁路管道,并对此条件下的过热器反暖操作和特殊工况下压力损失计算结果进行定性分析,确定了设计参数,优化了主蒸汽系统的工艺流程.

  13. 中国实验快堆全堆芯流量分配计算与试验%Calculation and Test of Core Flow Rate Distribution of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    刘一哲; 薛秀丽; 许义军; 冯预恒; 侯志峰

    2012-01-01

    Based on the core and primary circuit design of China Experimental Fast Reactor(CEFR), a multiple-channel thermal-hydraulic analysis code DAEMON was developed to calculate the core flow rate distribution and unsymmetric coefficient in different conditions. In the commissioning stage, a series of full-scale tests for reactor core were performed in CEFR with a permanent-magnet sodium flow meter. The numerical results of code DAEMON showed a good agreement with test data. The core hydraulic design was also validated with a view to the requirements of design criteria, commissioning and operation specifications.%针对中国实验快堆(CEFR)堆芯和一回路的设计特点,开发水力特性计算程序DAEMON,完成不同工况下的全堆芯流量分配计算,给出流量分配不均匀性等参数.在反应堆调试阶段,进行全堆芯流量分配试验.结果表明,程序计算值与试验值符合较好.在此基础上,验证了CEFR堆芯的流体力学设计,并为反应堆调试和运行提供了基础数据.

  14. 中国实验快堆全厂断电事故多维度热工耦合计算%Multi-dimension Coupled Simulation Method of Thermalhydralic Behavior in China Experimental Fast Reactor Under Blackout Accident

    Institute of Scientific and Technical Information of China (English)

    乔雪冬; 胡文军; 冯预恒; 张春明; 孙微; 赵守智

    2012-01-01

    多维度耦合方法是将传统的一维反应堆热工流体力学程序与三维流体动力学分析软件通过一定的耦合方法结合起来,实现反应堆局部复杂流体现象分析与系统计算的耦合方法.本工作根据中国实验快堆设计和运行经验,开发了基于Rubin和Fluent的耦合程序框架,完成了中国实验快堆全厂断电工况的计算和验证.计算结果表明,耦合方法对全场断电事故的计算结果合理可靠,是对一维系统程序分析方法的有益补充.%Multi-dimension coupled simulation is a method which combines the analysis of complex hydromechanical phenomenon in reactor with system calculation by the method of coupling traditional one-dimensional thermo hydrodynamic program with CFD software. The coupling frame was developed based on Rubin and Fluent codes. By the test calculation under the station blackout accident of China Experimental Fast Reactor (CEFR) , multi-dimension coupled simulation is proved reasonable and gives a efficient supplement to system calculation method.

  15. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

    The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3He, 6Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

  16. Experimental control of burn-up calculations for high temperature reactor fuel by introduction of a special alpha spectrometric method for the determination of transuranium content. An attempt to establish isotopic correlations

    International Nuclear Information System (INIS)

    In the field of high-temperature-reactor (HTR) fuel investigation there is a great interest in the experimental and calculational determination of heavy metal content under the aspects of burn-up physics and for the prediction of reliable data for reprocessing and waste management. Using a laser-micro-boring preparation method, high resolution alpha-spectroscopy and sophisticated computer decomposition programs we identify qualitatively and quantitatively most of the important actinide isotopes in irradiated HTR-fuel. Additionally we use data, delivered by gamma- and mass-spectroscopy of the same fuel samples. The evaluated results are compared with calculational results from the burn-up code ORIGEN, using a special generated HTR-neutron-cross-section library. In a first step we determine new cross sections for the uranium and plutonium isotopes depending on the irradiation conditions. In a second step we calculate correlations between the heavy metal isotopes and the burn-up or the fission products

  17. Siegel FIRST EXPERIMENTAL DISCOVERY of Granular-Giant-Magnetoresistance (G-GMR) DiagnosES/ED Wigner's-Disease/.../Spinodal-Decomposition in ``Super''Alloys Generic Endemic Extant in: Nuclear-Reactors/ Petrochemical-Plants/Jet/ Missile-Engines/...

    Science.gov (United States)

    Hoffman, Ace; Wigner-Weinberg, Eugene-Alvin; Siegel, Edward Carl-Ludwig Sidney; ORNL/Wigner/Weinberg/Siegel/Hollifeld/Yu/... Collaboration; ANL/Fermi/Wigner/Arrott/Weeks/Bader/Freeman/Sinha/Palazlotti/Nichols/Petersen/Rosner/Zimmer/... Collaboration; BNL/Chudahri/Damask/Dienes/Emery/Goldberg/Bak//Bari/Lofaro/... Collaboration; LLNL-LANL/Hecker/Tatro/Meara/Isbell/Wilkins/YFreund/Yudof/Dynes/Yang/... Collaboration; WestinKLouse/EPRI/PSEG/IAEA/ABB/Rickover/Nine/Carter/Starr/Stern/Hamilton/Richards/Lawes/OGrady/Izzo Collaboration

    2013-03-01

    Siegel[APS Shock-Physics Mtg., Chicago(11)] carbides solid-state chemistry[PSS (a)11,45(72); Semis. & Insuls. 5: 39,47,62 (79)], following: Weinberg-Siegel-Loretto-Hargraves-Savage-Westwood-Seitz-Overhauser-..., FIRST EXPERIMENTAL DISCOVERY of G-GMR[JMMM 7, 312(78); Google: ``If LEAKS Could KILL Ana Mayo''] identifIED/IES GENERIC ENDEMIC EXTANT domination of old/new (so mis-called) ``super''alloys': nuclear-reactors/spent-fuel-casks/refineries/jet/missile/rocket-engines in austenitic/FCC Ni/Fe/Co-based (so mis-called) ''super''alloys (182/82; Hastelloy-X,600,304/304L-Stainless-Steels,...,690!!!) GENERIC ENDEMIC EXTANT detrimental(synonyms!!!): THERMAL: Wigner's-disease(WD physics) [J.Appl.Phys.17,857(46)]/ Ostwald-ripening

  18. Fueling of tandem mirror reactors

    International Nuclear Information System (INIS)

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design

  19. 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

  20. 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

  1. Low cycle fatigue behaviour of neutron irradiated copper alloys at 250 and 350 deg. C. (ITER R and D Task no. T213)[International Thermonuclear Experimental Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singh, B.N.; Stubbins, J.F. [Illinois Univ., Dept. of Nuclear Engineering, Illinois (United States); Toft, P

    2000-03-15

    The fatigue behaviour of a dispersion strengthened and a precipitation hardened copper alloys was investigated with and without irradiation exposure. Fatigue specimens of these alloys were irradiated with fission neutrons in the DR-3 reactor at Risoe with a flux of {approx}2.5 x 10{sup 17} n/m{sup 2}s (E> 1 MeV) to influence levels of 1.0 - 1.5 x 10{sup 24} n/m{sup 2} (E> 1 MeV) at 250 and 350 deg. C. These irradiations were carried out in temperature controlled rigs where the irradiation temperature was monitored and controlled continuously throughout the whole irradiation experiment. Both unirradiated and irradiated specimens were fatigue tested in vacuum at the irradiation temperatures of 250 and 350 deg. C in a strain controlled mode with a loading frequency of 0.5Hz. Post-fatigue microstructures were examined using transmission electron microscopy and the fracture surfaces were investigated using scanning electron microscope. The present investigations demonstrated that the fatigue life decreases with increasing temperature and that the exposure to neutron irradiation causes further degradation in fatigue life at both temperatures. These results are discussed in terms of the observed post-fatigue microstructures and the fracture surface morphology. Finally, the main conclusions and their implications are summarised. (au)

  2. Mathematical modelling of water radiolysis kinetics under reactor conditions

    International Nuclear Information System (INIS)

    Experimental data on coolant radiolysis (RBMK-1000 reactor) were used to construct mathematical model of water radiolysis kinetics under reactor conditions. Good agreement of calculation results with the experiment is noted

  3. Fast breeder reactor research

    International Nuclear Information System (INIS)

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  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. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  6. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  7. Second-generation thermionic reactor (TRICE)

    International Nuclear Information System (INIS)

    During the past decade, advanced types of thermionic converters have been investigated and experimentally demonstrated. In addition to their substantially increased efficiency, these advanced (second-generation) converters also can condition the output power such as to remove many design constraints on the performance and complexity of the in-core thermionic reactor. A second-generation thermionic in-core reactor concept (TRICE) is described in which advanced-mode thermionic converters are integrated into a reactor core, and their output inductively-coupled to an external high-voltage AC electromagnetic output circuit. Because the electrical insulators and intercell gaps that are required in the flashlight type of in-core reactor are absent from the TRICE core, the size and complexity of the TRICE reactor are substantially less than for a flashlight type reactor of the same output power. The increased performance potentially available via the TRICE concept is estimated for a variety of reactor output power levels

  8. Isothermal calorimeter for reactor radiation dosimetry

    International Nuclear Information System (INIS)

    An isothermal calorimeter with thermistors for measuring absorbed dose rates from 104-5-6.105 rad/h in reactor experimental holes has been designed. A kinetics method for determining the equilibrium temperature difference has been developed, and its application in isothermal calorimetry proved. The expected accuracy in measurements within ± 2-5% has been proved by measurements carried out in the reactor. Some data obtained by measurements in the reactor RA are presented (author)

  9. 中国实验快堆典型钠阀温度分布研究%Study on temperature distribution in bellows seal sodium valve assembly of China Experimental Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    李生; 张东辉; 刘云焰

    2012-01-01

    The typical bellows seal sodium valves, as the important equipment in the sodium systems of China Experimental Fast Reactor, have a significant effect on the safety of fast reactors. The typical valves caused some problems in the test stage. The paper is to get the results of temperature distribution in the sodium valve assembly on stable heat transfer condition, using the Computational Fluid Dynamics (CFD) code. The paper also analyses the results computed under the condition of fixed thickness and the different height of thermal insulation materials. It gets a good conclusion through the comparison of measured and simulated results and the numerical simulation result is logical and meaningful.%中国实验快堆典型钠阀作为系统重要的涉钠设备,直接影响着反应堆系统的安全运行.中国实验快堆工程在调试和运行阶段面临着钠阀门带来的一系列问题.本文应用CFD软件计算了两种运行工况下典型钠阀稳态温度场分布,分析了保温层厚度一定、高度不同的情况下,钠阀门的温度场分布结果,并与实验结果进行了对比,证明结果是合理的有意义的.

  10. 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

  11. 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.

  12. Parametrization of the executive mechanisms control in the structure of the experimental plant of the IBR-2 and IBR-30 reactors

    International Nuclear Information System (INIS)

    CAMAC control block (BUIM) on the base of K1801VM1 microprocessor is intended for control of 32 executive mechanisms. Classification of the executive mechanisms by motor types, by displacement transducer types, by dynamic and other parameters is described. The access to control and monitoring information is distinguished by simplicity and reliability. The development of the BUIM microprogramm provides control of 32 executive mechanisms by turns. Parameter list is defined for each motor and may be extended. Computer software is invariant to executive mechanism parameters. All the control and monitoring problems are solved without computer. Functional buses are developed for reliability control of executive mechanisms of the experimental plant by corresponding adapters or directly. Eight motor types and three types of displacement transducers are connected to the control block. 18 refs.; 11 figs.; 1 tab

  13. 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)

  14. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  15. 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)

  16. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  17. 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

    International Nuclear Information System (INIS)

    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

  18. 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

  19. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  20. Experimental study on tar destruction in a two stage fixed-bed reactor%两段式固定床反应器中焦油脱除的实验研究

    Institute of Scientific and Technical Information of China (English)

    吴文广; 罗永浩; 陈祎; 苏毅; 陈亮; 王芸

    2012-01-01

    通过两段式固定床反应器实验,研究了热裂解、部分氧化和炭层转化三种方法对焦油脱除的效果,并研究了生物质种类、反应温度、停留时间、生物质焦的粒径及种类等因素对热解焦油的脱除和转化规律.结果表明,随着温度的升高,三种脱除方法中焦油生成量下降,且降幅逐渐减小,实验过程中无论采取何种方法,都难以将焦油完全脱除;部分氧化和炭层转化对焦油的脱除效果都较相同温度条件下的热裂解要好,且在焦油脱除效果上,炭层转化>部分氧化>热裂解;联合部分氧化和炭层转化可达最高的焦油脱除效率,三种生物质热解焦油经1000℃联合脱除后产量分别为,稻秆0.43%、玉米秆0.61%和杉木屑1.15%,转化率分别达到98.28%、97.23%和96.29%;相同实验条件下稻秆的热解焦油最容易脱除,这与其物料中含氧量较高有关;生物质焦种类对焦油的脱除效果影响较小.%Methods of thermal cracking, partial oxidation and char bed conversion on tar destruction has been investigated by a two stage fixed-bed reactor, effects of fuel type, temperature, residence time, char particle size and char type on tar destruction are considered. The result indicates that tar conversion efficiency increase with the second stage reactor temperature increasing in all three kinds of conversion methods. Partial oxidation and char bed conversion is more effective in tar destruction compared to thermal cracking. Associated with partial oxidation and thermal cracking, char bed can get least tar yield. Three kinds of biomass tar yield in the experimental condition of 1 000℃ is: rice straw 0.43% , corn straw 0. 61% and fir sawdust 1. 15% , and the corresponding tar conversion efficiency is 98.28% , 97. 23% and 96.29% respectively. Tar yield content of each conversion methods are decreasing with reactor temperature increase. It is really difficult to removal all tar