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Sample records for braunschweig experimental reactor

  1. Experimental and Numerical Investigations on the Operation of the Hypersonic Ludwieg Tube Braunschweig

    Science.gov (United States)

    Estorf, M.; Wolf, T.; Radespiel, R.

    2005-02-01

    In summer 2003 the Hypersonic Ludwieg Tube Braunschweig (HLB) has been commissioned. It has been designed for Machnumber M = 6. The operative range of the unit Reynoldsnumber is between 3 · 106 1/m and 20 · 106 1/m. The test section has 500 mm diameter and the run time with near steady flow conditions is 80 ms. First measurements of the pitot pressure within the test-section have shown good transverse uniformity, but measurements of the total temperature within the storage tube revealed a strong stratification accompanied by convective flow within the heated section of the tube. In this paper we present recent measures to attenuate the stratification. Measurements of the temperature distribution within the testsection have been performed. Further we compare first results of invisvid numerical simulations of the unsteady onset of flow in the HLB to measured data. Key words: hypersonic flow, Ludwieg tube, flow measurement techniques, numerical flow simulation.

  2. Irrigation of treated wastewater in Braunschweig, Germany

    DEFF Research Database (Denmark)

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

    2007-01-01

    In this study the fate of pharmaceuticals and personal care products which are irrigated on arable land with treated municipal waste-water was investigated. In Braunschweig, Germany, wastewater has been irrigated continuously for more than 45 years. In the winter time only the effluent of the...

  3. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

    JAEA cooperation in remote monitoring focuses on the Joyo Experimental Reactor at the O'arai Research and Development Center. Joyo performs irradiation of test fuels to support development of the fast reactor cycle in Japan, both in international cooperation and in support of the Monju fast reactor, which is now undergoing reconstruction. The tour included an introduction at the model, a visit to the control room, entry into the containment vessel, and viewing of remote monitoring equipment in the Fresh Fuel Storage and at one of the Spent Fuel Ponds. (author)

  4. Report on the operation of the Forschungs- und Messreaktor Braunschweig (FMRB) for the year 1982

    International Nuclear Information System (INIS)

    In 1982 the Forschungs- und Messreaktor Braunschweig (FMRB) has been in action for 2618 hours without any serious disturbance. The released power for this period amounts to 2507 MWh. The experience in the field of reactor operation, in the radiation protection work as well as in the experiments at the beam tubes is reported. (orig.)

  5. Report on work on the Forschungs- und Messreaktor Braunschweig (FMRB) for the year 1977

    International Nuclear Information System (INIS)

    In 1977 the Forschungs- und Messreaktor Braunschweig (FMRB) has been in action for 3289 hours without any serious disturbance. The experience in the reactor operation, in the radiation protection work as well as in the experiments at the beam tubes is reported. (orig.) 891 UA

  6. 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

  7. China experimental fast reactor

    International Nuclear Information System (INIS)

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*1015 n/cm2/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  8. 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

  9. 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

  10. Experimental reactor physics

    International Nuclear Information System (INIS)

    Neutronic experiments in moderators, subcritical assemblies, critical assemblies, and nuclear reactors are described, as well as the techniques of radiation measurements necessary to perform these experiments. Previously dispersed data from government reports, journal articles, and specialized monographs are codified. Original information drawn from the author's experience is included, especially on the pulsed source technique, spectrum measurements, research reactors, and exponential assemblies. The book provides the essential information for carrying out, analyzing, and understanding the experiments. Theory is kept to a minimum. Emphasis is placed on the physics of the situation, and the importance of estimating error as well as the mean value of a measured quantity

  11. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  12. 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.

  13. Status of Fusion Experimental Reactor (FER) design

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI in line with a long-range plan for fusion reactor development laid out in the long-term program of the Atomic Energy Commission issued in 1982. The FER succeeding the tokamak device JT-60 is a tokamak reactor with a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. The paper describes recent developments of the FER design concept

  14. 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. ...

  15. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  16. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

  17. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

  18. 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

  19. 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

  20. 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

  1. ETRR-2 nuclear reactor: Experimental results

    International Nuclear Information System (INIS)

    The report describes the experimental results from a neutronic and thermalhydraulic point of view. The thermalhydraulic experiments included are steady state, loss of flow transient and negative reactivity insertion. The neutronic experiments given are critical configurations, control rod calibrations and second shutdown system reactivities. The goal of the report is to provide sufficient experimental details to enable simulation of the experiments. It should be used in conjunction with the companion report, ETRR-2 Nuclear Reactor: Facility Specification. (author)

  2. 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)

  3. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  4. Safety scenario for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A scenario to ensure the safety of the Fusion Experimental Reactor (FER) is proposed. The safety features of a fusion reactor are given and their impacts on the safety design are shown. The requirements in the design of major components of FER to achieve safety and the safety evaluation process are described. The results of the evaluation showed that even in the event of the maximum credible accidents, the radiological consequence to the public can be held at an acceptable level. The applicability to FER of various aspects of the regulations for facilities treating large amounts of radioisotopes is discussed with a positive conclusion. (author). 11 refs, 1 fig

  5. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-01

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

  6. China experimental fast reactor; Le reacteur rapide experimental chinois

    Energy Technology Data Exchange (ETDEWEB)

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)

    2007-07-15

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  7. The International Thermonuclear Experimental Reactor configuration evolution

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

  8. 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)

  9. 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

  10. Shielding design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This report first describes the basic design philosophy of radiation shields for the fusion experimental reactor (FER) which has been proposed to be the next step machine to JT-60. Next, geometrical models and calculation parameters for shielding calculations were investigated to establish the standard design calculation methods, and accuracy of the calculation was evaluated. Further, irradiation properties of in-vessel components and bulk shielding properties were summarized in the useful form for the future design works. (author)

  11. Modeling a nuclear reactor for experimental purposes

    International Nuclear Information System (INIS)

    The Loss-of-Fluid Test (LOFT) Facility is a scale model of a commercial PWR and is as fully functional and operational as the generic commercial counterpart. LOFT was designed and built for experimental purposes as part of the overall NRC reactor safety research program. The purpose of LOFT is to assess the capability of reactor safety systems to perform their intended functions during occurrences of off-normal conditions in a commercial nuclear reactor. Off-normal conditions arising from large and small break loss-of-coolant accidents (LOCA), operational transients, and anticipated transients without scram (ATWS) were to be investigated. This paper describes the LOFT model of the generic PWR and summarizes the experiments that have been conducted in the context of the significant findings involving the complex transient thermal-hydraulics and the consequent effects on the commercial reactor analytical licensing techniques. Through these techniques the validity of the LOFT model as a scaled counterpart of the generic PWR is shown

  12. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    The Fusion Experimental Reactor (FER) being developed at JAERI as a next generation tokamak to JT-60 has a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. During FY82 and FY83 a comprehensive and intensive conceptual design study has been conducted for a pulsed operation FER as a reference option which employs a conventional inductive current drive and a double-null divertor. In parallel with the reference design, studies have been carried out to evaluate advanced reactor concepts such as quasi-steady state operation and steady state operation based on RF current drive and pumped limiter, and comparative studies for single-null divertor/pumped limiter. This report presents major results obtained primarily from FY83 design studies, while the results of FY82 design studies are described in previous references (JAERI-M 83-213--216). (author)

  13. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (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. Conceptual design of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been performed. The FER has an objective of achieving selfignition and demonstrating engineering feasibility as a next generation tokamak to JT-60. Various concepts of the FER have been considered. The reference design is based on a double-null divertor. Optional design studies with some attractive features based on advanced concepts such as pumped limiter and RF current drive have been carried out. Key design parameters are; fusion power of 440 MW, average neutron wall loading of 1MW/m2, major radius of 5.5m, plasma minor radius of 1.1m, plasma elongation of 1.5, plasma current of 5.3MA, toroidal beta of 4%, toroidal field on plasma axis of 5.7T and tritium breeding ratio of above unity

  16. Research Reactor Benchmarking Database: Facility Specification and Experimental Data

    International Nuclear Information System (INIS)

    This web publication contains the facility specifications, experiment descriptions, and corresponding experimental data for nine different research reactors covering a wide range of research reactor types, power levels and experimental configurations. Each data set was prepared in order to serve as a stand-alone resource of well documented experimental data, which can subsequently be used in benchmarking and validation of the neutronic and thermal-hydraulic computational methods and tools employed for improved utilization, operation and safety analysis of research reactors

  17. 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

  18. Experimental investigation of hydraulic characteristics of tank reactor model

    International Nuclear Information System (INIS)

    Experiments for studying the hydraulic characteristics of a vessel reactor model at the MR stand described. The hydraulic model of a two-loop reactor of the vessel type is described. The experimental data are obtained in the wide range of the stand operating parameters, including the emergency modes of the reactor model operation with the total shut-down of one feed pump

  19. The jules Horowitz reactor: experimental possibilities

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor is a new research reactor dedicated to materials and nuclear fuel testing. This reactor will be located at the CEA research centre at Cadarache and the first criticality is foreseen around 2008-2010. After a short overview of the fuel, materials testing and service possibilities of the reactor this paper is reviews two techniques which are intended to be developed on the Jules Horowitz Reactor to measure fission gas release: - The development of a new sensor to be used on-line and to measure pressure and composition of the gas released in a rod or elsewhere. - The Fission Product Laboratory. (author)

  20. Experimental results from the BNL zero power reactor HITREX

    Energy Technology Data Exchange (ETDEWEB)

    Kitching, S.J.; Lewis, T.A.; Playle, T.S.

    1973-10-15

    This report presents experimental results obtained with the BNL reactor Hitrex. Measurements of reactivity, and of thermal and fast neutron reaction rate distributions have been made with various experimental control rod configurations.

  1. Impurity control in Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Poloidal divertor system is employed as the impurity control measure in Fusion Experimental Reactor (FER). The authors will report the overall survey of impurity control physics in FER. The results obtained are as follows. (1) The triple-valued solutions of divertor plasma equilibrium are obtained as a function of incoming ion flux. Engineering design is carried out based on the stable dense and cold divertor plasma. (2) Low density and high temperature solution disappears when the geometry is extremely closed (chamber length=50cm and void width=O). (3) Plasma temperature can become slightly high on the side of exhaust duct near the plate. (4) Erosion rate on the first wall by charge-exchange neutrals is recognized to be about 1cm/year by DEGAS code that is also obtained by even our divertor code and simple order estimation. (5) Cold and dense divertor plasma could be formed during noninductive current drive phase either LHRF or NBI, if the current drive efficiency is improved

  2. Fusion Experimental Reactor (FER) design concept

    International Nuclear Information System (INIS)

    Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI within the frame of the longterm program of the Atomic Energy Commission issued in 1982. The major mission of the FER, which is the device planned to succeed the JT-60 tokamak device, is realizing a self-ignited, long burn D-T plasma and demonstrating engineering feasibility of fusion energy. The reference design concept is based on a quasi-steady state operation scenario. This scenario includes non-inductive current drive at low plasma density during startup and power dwell period and conventional inductive current drive for high density plasma burn periods (2000 s). Key design features and parameters are as follows: non-breeding blanket (shield blanket); single-null poloidal divertor; lifetime neutron fluence of 0.3 MW·y/m2; fusion power of 300 MW; average wall loading of 0.68 MW/m2; plasma major radius of 5.2 m; plasma minor radius of 1.12 m with the elongation of 1.5; and plasma current of 5.9 MA. (author). 4 refs, 10 figs, 9 tabs

  3. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li2O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

  4. 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)

  5. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

  6. Experimental possibilities of IGR reactor for the researches on the nuclear reactor safety

    International Nuclear Information System (INIS)

    The IGR reactor (National nuclear centre of the Republic of Kazakstan, Kurchatov) with high technical and neutron-physical properties has wide experimental possibilities for the dynamic studies. On this reactor possible curried out two general types of regimes. First regime is a 'flare', non-regular neutron impulse of power of bell form. In this regime the maximum flux density of thermal neutrons. Second regime is a 'impulse', regulated on the given regime (law) power impulse. Profile of power change in this regime has sections of linear ascent and fall, sections of stationary power. IGR reactor has pneumatic hydraulic stand, provided accumulation in the ramps of high pressure. Experimental volume of the reactor are composites central and lateral channels, which are passed through active zone of height equal to 1400 mm. The above mentioned possibilities of IGR reactor are provided unique conditions for studies in the field of nuclear reactor safety

  7. 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.

  8. 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.

  9. Tajoura reactor conversion and preliminary neutronic experimental data

    International Nuclear Information System (INIS)

    Reactor LEU fuel conversion means that a certain procedures must be done according to this process. This paper presents the local experience of this work and focuses on the experimental data of neutron flux measurements along the fuel and reflector cells of Tajoura research reactor immediately followed the reactor building up critical mass. The study was done by applying the neutron activation technique with the irradiation of thin gold foils along the selected core cells and the results shows agreement with that publications based on the comparison between low and high enrichment fuels neutron fluxes. (author)

  10. Tajoura reactor conversion and preliminary neutronic experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Aburwes, M.E.N. [Reactor Department, Nuclear measurements lab, Renewable Energies And Water Desalinaton Research Center, Tajoura (Libyan Arab Jamahiriya)

    2008-10-29

    Reactor LEU fuel conversion means that a certain procedures must be done according to this process. This paper presents the local experience of this work and focuses on the experimental data of neutron flux measurements along the fuel and reflector cells of Tajoura research reactor immediately followed the reactor building up critical mass. The study was done by applying the neutron activation technique with the irradiation of thin gold foils along the selected core cells and the results shows agreement with that publications based on the comparison between low and high enrichment fuels neutron fluxes. (author)

  11. Architecture of the ETR [experimental test reactor] systems code

    International Nuclear Information System (INIS)

    TETRA, a tokamak systems code capable of modeling experimental test reactors (ETRs), was developed in a joint effort by participants of the fusion community. The first version of this code was constructed to model devices similar to the Tokamak Ignition/Burn Engineering Reactor (TIBER) in configuration and design. A major feature of this code is its ability to perform optimization studies. Future work will include broadening the scope of the code, particularly in the area of materials selection, to more accurately simulate tokamak configurations such as the Next European Torus (NET) and the Fusion Engineering Reactor (FER). 18 refs., 2 figs., 4 tabs

  12. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

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

  14. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    The SCARABEE in-pile experimental programme comprised a series of tests on unirradiated fuel pins, either single or in seven-pin clusters. The main objective was to obtain information on the mode and consequences of fast reactor fuel pin failure in conditions representative of loss of cooling in a LMFBR. The application of such programmes in full scale reactors leads to the great importance of the interpretation of experimental observations. The interpretation of that programme was carried out jointly by CEA, KFK and UKAEA; this international collaboration led to a sharper focusing on essential features to be modelled in experiments and computer codes and to a valuable convergence of views

  15. Design features of BREST reactors and experimental work to advance the concept of BREST reactors

    International Nuclear Information System (INIS)

    Principle design features of BREST-300 (300 MWe) and BREST-1200 (1200 MWe) lead.cooled fast reactors are presented in this paper. Several experimental works have been performed or under way in order to justify lead-cooled reactor design concepts. BREST reactor designs of different outputs have been developed using the same principles. In conjunction with the increased output and the implement of inherent safety concept, a number of new solutions, which may be applied to the BREST-300 reactor design too, have been considered in the BREST-1200 reactor design. The new design features adopted in the BREST-1200 reactor design include: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by--pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

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

    International Nuclear Information System (INIS)

    This report describes the results of the reactor configuration/structure design for the fusion experimental reactor (FER) performed in FY 1986. The design was intended to meet the physical and engineering mission of the next step device which was decided by the subcommittee on the next step device of the nuclear fusion council. The objectives of the design study in FY 1986 are to advance and optimize the design concept of the last year because the recommendation of the subcommittee was basically the same as the design philosophy of the last year. Six candidate reactor configurations which correspond to options C ∼ D presented by the subcommittee were extensively examined. Consequently, ACS reactor (Advanced Option-C with Single Null Divertor) was selected as the reference configuration from viewpoints of technical risks and cost performance. Regarding the reactor structure, the following items were investigated intensively: minimization of reactor size, protection of first wall against plasma disruption, simplification of shield structure, reactor configuration which enables optimum arrangement of poloidal field coils. (author)

  17. Experimental research of reactor core flooding

    International Nuclear Information System (INIS)

    The results are presented of experiments performed with the aim of finding the influence of the method of fixing the thermocouples for measuring the distribution of temperature to the wall of fuel pin simulator. This influence was found for the purpose of emergency core flooding. First experimental results on the effect of nitrogen dissolved in the water on the velocity of the cooling wave are given. These experiments were carried out under the following conditions: initial temperature in pin centre 300 to 600 degC, velocity of water at the inlet into the measuring section 3.5 to 20 cm/s, and atmospheric pressure in the model. (author)

  18. 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)

  19. ITER [International Thermonuclear Experimental Reactor] physics basis

    International Nuclear Information System (INIS)

    The physics basis of ITER has been developed from an assessment of the present knowledge of tokamak physics with allowance for improvements in that knowledge during the design and construction phases of ITER. The assessment has been carried out by the ITER design team in collaboration with the international fusion program, including participation by the experimental teams of all of the major toroidal experiments. The physics basis consists of guidelines for energy confinement, operational limits, power and particle control, disruptions, current drive and heating, alpha particle physics, and plasma control. The ITER physics group has worked with the engineering groups to implement these guidelines. In addition, a preliminary design for the plasma diagnostics for ITER has been developed and an operational program has been planned. In many cases, the physics issues have not been fully resolved, and a physics R ampersand D program has been developed to complete the physics basis for ITER. 16 refs., 3 tabs

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

    International Nuclear Information System (INIS)

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

  1. 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 ...

  2. 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).

  3. 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)

  4. Conceptual design study of Fusion Experimental Reactor (FY87FER)

    International Nuclear Information System (INIS)

    This report describes the results of the design coordination and the conceptual design study on plant systems which have been carried out as a part of the design work for the Fusion Experimental Reactor (FY87 FER). The former contains the selection of the reference concept for FY87 FER and giving it flexibility, directions for study and assessment of low physics risk reactors, and the revision of system integration, while the latter mainly describes the design philosophies, construction of systems, and the results of designs and analyses of processes and systems. (author)

  5. Maintenance of torus components for the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Maintenance of torus components is one of key technologies for the Fusion Experimental Reactor (FER), which is the device planned to succeed the JT-60 tokamak device. The objective of the present study is to develop a reliable, feasible and simple maintenance systems for torus components such as divertor and movable shield modules. This paper describes the reactor structure and its maintenance scheme of FER and the maintenance systems for torus components. A 1/4-scale mock-up of FER is also introduced, which was made to demonstrate the feasibility of the maintenance system for torus components of FER

  6. Conceptual design study of Fusion Experimental Reactor (FY87FER)

    International Nuclear Information System (INIS)

    The design study of Fusion Experimental Reactor(FER) which has been proposed to be the next step fusion device has been conducted by JAERI Reactor System Laboratory since 1982 and by FER design team since 1984. This is the final report of the FER design team program and describes the results obtained in FY1987 (partially in FY1986) activities. The contents of this report consist of the reference design which is based on the guideline in FY1986 by the Subcomitees set up in Nuclear Fusion Council of Atomic Energy Commission of Japan, the Low-Physics-Risk reactor design for achieving physics mission more reliably and the system study of FER design candidates including above two designs. (author)

  7. 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)

  8. 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.

  9. Remote maintenance design for Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Design of Fusion Experimental Reactor, FER, has been conducted by Japan Atomic Energy Research Institute (JAERI) since 1981. Two typical reactors can be classified in general from the viewpoints of remote maintenance among four design concepts of FER. In the case of the type 1 FER, the torus module consists of shield structure and blanket, and the connective joints between toruses provided at the outer region of the reactor. As for the type 2 FER, the shield structure is joined with the vacuum cryostat, and only the blanket module is allowed to move, but connection between toruses are located in the inner region of the reactor. Comparing type 1 with type 2 FER, this paper describes on the remote maintenance of FER including reactor configurations, work procedures, remote systems/equipments, repairing facility and future R and D problems. Reviewing design studies and investigation for the existing robotics technologies, R and D for FER remote maintenance technology should be performed under the reasonable long-term program. The main items of remote technology required to start urgently are multi-purpose manipulator system with performance of dextrousity, tele-viewing system which reduces operator fatigue and remote tests for commercially available components

  10. 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

  11. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  12. 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)

  13. 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...

  14. 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...

  15. Design Features and Operating Experience of Experimental Fast Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The IAEA has begun an initiative to help coordinate Member State efforts in the field of fast neutron nuclear reactors. This initiative is primarily targeted at the preservation of knowledge in the areas of design, construction and operation, for both experimental and power fast reactors. The ultimate goal of this activity is to establish a comprehensive, international inventory of fast reactor data and knowledge, which will be an essential resource for the future development and deployment of fast reactor technology. In this project, carried out within the framework of the

  16. 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.

  17. 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

  18. 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

  19. 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)

  20. 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.

  1. Evaluation of skyshine calculation method for fusion reactor and application to fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    In the design of the reactor room for a fusion reactor, the cost of the room strongly depends on the thickness of the roof because the area of the roof is generally large. The roof thickness is mostly determined by the requirement to reduce the skyshine dose rate level at the site boundary below the assigned value. Therefore the accurate evaluation of the skyshine dose becomes important for the design of the reactor room. Skyshine dose for a D-T fusion reactor has been evaluated by a number of researchers but the agreement is not so good. In this report, the first collision source is used with two-dimensional SN transport method to form DOT3.5-GRTUNCL-DOT3.5 coupled calculation flow. The validity of the methodology was first shown by calculating the skyshine dose from a 14 MeV neutron source and comparing the calculated results with the measured results. This methodology was then used to calculate the skyshine dose for the Fusion Experimental Reactor (FER). The calculated results were compared with those from several other methods to clarify the mutual difference. (author)

  2. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 1015 n/cm2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author)

  3. 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.

  4. Experience with EBR-II [Experimental Breeder Reactor] driver fuel

    International Nuclear Information System (INIS)

    The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

  5. 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.

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

    International Nuclear Information System (INIS)

    This report describes the results of applicability studies for the negative ion-based neutral beam injector to the Fusion Experimental Reactor (FER). The operation scenario of FER has been proposed to adopt the neutral injection method as one of candidates, which has three functions of heating, current drive and profile control. One of the fundamental requirements is the tangential injection of the neutral beam. For neutral beam injectors, three port sections are available. Supposing to adopt the beam line with the straight long neutralizer which has been designed in JAERI, the geometrical arrangement was determined so as to avoid any trouble to the reactor structure. The conceptual study for major components which compose the beam line system was carried out including the estimation of the neutron streaming. The power supply system was studied also and the work was concentrated on the acceleration power supply which requires the output voltage of 500 kV and fast cut-off time. A basic concept, in which a inverter with a AC switch is used and the frequency of the supplied AC line is increased was proposed. In these works, the configuration of the neutral beam injection system was detailed and it was shown that the beam line seems to be well implemented with the geometrical constraints related to the reactor configuration. (author)

  7. 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.)

  8. Tasks related to increase of RA reactor exploitation and experimental potential, 03. Crane for handling the vertical experimental channels of the RA reactor - design project

    International Nuclear Information System (INIS)

    Within the work related to improvement of experimental potential of the RA reactor, this document describes the design project of the new crane for handling the vertical experimental channels of the RA reactor, engineering drawings of the crane main elements, mechanical part, design project of the electrical part of the crane and cost estimation

  9. 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

  10. 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

  11. Radiochemical characterisation of graphite from Juelich experimental reactor (AVR)

    International Nuclear Information System (INIS)

    Graphite built-in nuclear reactors may receive a high neutron dose for a long period. Depending on its chemical composition a lot of activation products are produced. In addition, there is more or less fission product contamination depending on the location. The migration of fission products may be supported by high temperatures which occur in high temperature reactors. At the Juelich 15 MWe High Temperature Gas-cooled experimental Reactor AVR (Arbeitsgemeinschaft Versuchsreaktor) two different types of nuclear graphite had been in use. High-purity graphite was used as basic material for core structures of the AVR. Insulation layers from carbon bricks (graphite with larger amounts of impurities) surrounding the graphite reflector were used to protect the metallic structures from high temperatures. For many reasons it is important to know the amount of contamination of graphite and carbon bricks with activation products and fission products. The head end of nuclear graphite analytics must be the incineration. Volatile activities (14C, 3H, 36Cl ...) must be caught for determination. In case of handling dustlike samples the incineration furnace must be small enough to be operated in a glove box. The resulting ashes can be used for determining all non volatile nuclides with different radiochemical methods. In early 1999 some graphite and carbon brick samples from AVR-reactor had been taken by drilling. The samples had been analysed in our laboratories at Juelich research centre. For incineration we used a vertical quartz-tube which dips at the bottom into a small electric furnace. Tritium, 14C and 36Cl are caught in washing bottles. After further preparation, they are determined by LSC. After dissolving the ashes, the elements are separated by ion exchange, extraction methods and HPLC. The radionuclides are then determined by a-spectrometry, LSC, low level g-spectrometry and x-ray spectrometry. (author)

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. In-pipe experimental needs for resolution of principal reactor safety issues in commercialization of fast reactors

    International Nuclear Information System (INIS)

    This paper describes major research subjects and approaches on reactor safety for commercialization of fast reactors including recriticality issue in core disruptive accident sequences. To achieve the research objective for these major subjects, a new in-pile safety experimental program named SERAPH (safety engineering reactor for accident phenomenology) is proposed, and the conditions of the proposed tests and the major requirements for the facility are formulated. 13 refs., 5 figs., 1 tab

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

    International Nuclear Information System (INIS)

    This report describes the results of a conceptual study on the RF system in the typical candidates for the Fusion Experimental Reactor (FER), which were picked out through the '86FER scoping studies. According to the FER operation scenario, three RF systems, that is, ICRF (heating), LHRF (current drive and heating), ECRF (auxiliary heating) were studied. Main concern in these RF systems is the launcher, which may be so designed that required power match the geometrical constraints of the reactor. Then studies were concentrated on the launcher configuration. A prug-in concept of the launcher was adopted in each system and vacancies except transmission space were filled with water. The ICRF launcher had the 2 x 2 loop arrays antenna and the faraday shield area of 1.5 m x 1 m to provide a power of 20 MW. The LHRF launcher had the grillantenna with 28 x 8 open waveguides, and included multi junction-type power splitters which were connected to 56 transmission wave guides. The grild was designed to have two functions of current drive and heating, and provide a power of 20 MW each. The ECRF launcher had a boundle of open wave guides which a reflection mirror each, and three plain mirrors. Assuming a oscillator unit size of 200 kW, it had 40 oversized wave guides to provide a power of 3 MW. (author)

  18. Summary of 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. Starting from 1984 FER design is now being reviewed and redesigned. This report is the summary of the report which describes the results obtained in the review and redesign activities in 1984. The following three steps are followed in those activities ; critical issues study step in which FER critical issues were reviewed and the frame of FER design was revised, torus structure selection step in which a few options within the frame for FER were examined and design step in which major components of the torus structure were designed. 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)

  19. Conceptual design study of the fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    A conceptual design study of the Fusion Experimental Reactor (FER) is presented. FER is planned, on the basis of a domestic programme, as a device to explore reasonable minimum physics and technological issues necessary to proceed to DEMO. Among various concepts, including the improvement of ITER-like design, the reference design of FER is chosen as size-minimum and a detailed design study is performed. LHCD assist and a single null divertor configuration are employed to reduce the device size. Simplification and the resultant reliability are attained by the appropriate choice of the fluence, 0.3 MW·a·m-2. A variety of new ideas are explored to develop the FER concept. A layered structure of the divertor for a reliable maintenance scheme and a uniform vacuum vessel with a thin double wall structure for providing structural simplicity and tritium double containment are typical examples. R and D programmes for these key reactor components are now actively being promoted in conjunction with this design activity. (author). 4 refs, 2 figs, 1 tab

  20. 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.

  1. Study on plasma ignition of JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Heating the plasma in JAERI Experimental Fusion Reactor up to the equilibrium operating state has been studied with a time dependent zero-dimensional model. The neoclassical or pseudoclassical scaling-law plays a leading part of the plasma diffusion in the low temperature region below several keV and the trapped-ion scaling-law does so in the higher region. The plasma temperature is raised to 1 keV by 10 sec Joule-heating. The plasma is heated up to the equilibrium operating state of plasma temperature 7 keV and electron density 1.1 x 1020 m-3 by 10 sec neutral beam injection heating with injection power 28 MW and fueling rate 3 x 1019 m-3s-1. (auth.)

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

    International Nuclear Information System (INIS)

    This report describes the study on safety for FER(Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. This report consists of two chapters. The first chapter of this report summaries the FER system and describes FMEA(Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including the purification, isotope separation system and storage system. Here, probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA. (author)

  3. 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)

  4. Experimental estimation of the neutron flux density at the reconstructed Rossendorf research reactor

    International Nuclear Information System (INIS)

    The Rossendorf Research Reactor was reconstructed in the years 1986-1989. During start up of the reactor the neutron flux density was investigated in the reactor core and the outer irradiation channels by the help of activation probes and self-powered neutron detectors. The report includes the most important experimental results and a brief description of the measuring techniques. (orig.)

  5. Experimental and analytic investigation of the ITU TRIGA Mark-II reactor core

    International Nuclear Information System (INIS)

    Experimental and analytical studies have been performed to determine the temperature distribution as a function of reactor power in the TRIGA Mark-II reactor at the Istanbul Technical University (ITU). The lumped parameter model with four governing equations was used in the analytical model. Based on the mathematical model, a computer code has been developed for calculating fuel and coolant temperatures in the reactor core. The calculated results for fuel and coolant temperature in the reactor core for different reactor power levels have been compared with the experimental data. Agreements between experiment and results from the computer code are fairly good. (orig.)

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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

  12. Design of separated first wall for fusion experimental reactor

    International Nuclear Information System (INIS)

    Design studies and R and D activities on the separate first wall for tokamak fusion experimental reactors have been progressed at JAERI. The first wall has a high probability of unexpected damage because of the uncertainties in local heat and particle loads and it requires easy replacement in case of failure. In order to satisfy the requirement of assembly and maintenance, the first wall mechanically separated and separately cooled from a massive blanket module has been proposed as a promising concept with a number of advantageous features, such as easy handling during assembly/disassembly due to light weight (∝350 kg), short down-time for maintenance operation, minimized amount of radwaste and so on. A fail-safe structure, which is consistent with in-service-inspection requirements, has been realized by employing a reliable double-walled thin shell structure sandwiching metal mesh. A quilting structure of austenitic stainless steel (SS316) cooled by low pressure (2 MPa), low temperature (100-150 C) water is employed to accommodate high surface heat flux of more than 0.3 MW m-2 and nuclear heating together with large electromagnetic loads up to 2 MPa. This paper describes the outlines of the structural design of the separated first wall, cooling and manifolds, mechanical connection to blanket structure, fabrication procedure, results of thermo-mechanical analyses and related R and D activities performed at JAERI. (orig.)

  13. 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)

  14. 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.

  15. Experimental reactors in the European Community and their utilization

    International Nuclear Information System (INIS)

    Research and test reactors which in the first years of the peaceful use of nuclear energy had to found the basis for building and operation of commercial nuclear power plants, having achieved their aim, have faded into the background of the report. They still play an important role, however, for the further development of today's power reactor generation and for the development of progressing reactor lines as well as for fuel and material irradiation, for isotope production and, last but not least, for research and training. At the moment, over 100 test reactors are being operated in the widest sense in the European Community. In the present survey, their purpose and charge are dealt with particular consideration to the more important materials test reactors and to the programme reactors. (orig./LH)

  16. 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

  17. Experimental method for reactor-noise measurements of effective beta

    International Nuclear Information System (INIS)

    A variance-to-mean noise technique, modified to eliminate systematic errors from drifting of reactor power, has been used to infer integral values of effective beta for uranium and plutonium fueled fast reactor modk-ups. The measurement technique, including corrections for a finite detector-electrometer time response, is described together with preliminary beta measurement results

  18. 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.)

  19. Experimental application of rhodium detectors in control systems of nuclear reactors

    International Nuclear Information System (INIS)

    The theoretical basis for construction of correcting devices for eliminating the inertness of rhodium detectors is examined. The experimental application of an improved rhodium detector in the IRT-2000 Sofia reactor is described. The flow chart of its inclusion in the reactor core and connection to the control circuit is given. The results confirm the concept that the local rhodium detectors with corrected inertness can be used for control of the reactor capacity and quick-acting safety of the reactor core of WWER type reactors. 3 figs., 4 refs

  20. 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.

  1. 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.)

  2. The TRIGA reactor Frankfurt construction and experimental facilities

    International Nuclear Information System (INIS)

    The new reactor FRF 2 was designed by Gutehoffnungshutte Sterkrade AG in cooperation with the reactor group of the Institut fur Kernphysik. The maximum power level is 1 MW; later installation of facilities for pulsed operation is possible. Performance and design data of the FRF 2 are given. The reactor is expected to start operation in 1973. Since the FRF 2 will be installed inside the biological shield and reflector of the FRF 1, the FRF 2 core has to correspond to the FRF 1 core structure

  3. Experimental capabilities of the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service

  4. Development of facilities to irradiate materials in the RA1 and RA3 experimental reactors

    International Nuclear Information System (INIS)

    To study the properties of the materials under irradiation, devices and facilities were designed to work at experimental reactors of National Atomic Energy Commission. The radiological protection of the operators and the influence of the irradiated materials on the radiological inventory of the reactors were the most important aspects considered during the design stage. In the present work devices to operate in the argentine reactor 'Reactor Argentino (RA)', RA1 and RA3 experimental reactors are shown. These devices are dedicated to the study of the radiation damage by measuring property changes related to dimensional integrity and embrittlement of materials in zirconium alloys, steels and other materials used in nuclear reactors. The emphasis is on the previsions adopted to minimize the activation of their components and the criteria applied to guarantee the safety of the operators during their performance and after their subsequent dismantling. (author)

  5. 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...

  6. 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.

  7. Possibilities of Kazakhstan experimental base for space nuclear reactors elements testing

    International Nuclear Information System (INIS)

    To the mid of 70-th in Kazakhstan the surface developing base for space nuclear reactors elements testing was created. The base consists of three test complexes. Two of them - the complexes of test reactors 'Baikal-1' and IGR - are situating on the Semipalatinsk test site, and the third one - complex of WWR-K research reactor - is situating in Alatau village nearby to Almaty city. On 'Baikal-1' and IGR complexes the testings for fuel elements, fuel assemblies, modules and prototypes of nuclear rocket engine reactor and nuclear energetic engine units with turbine-engine energy transmission on the base solid-phase reactor were carrying out. On the WWR-K reactor complex the testing of power generating channels of thermal-emission transmission reactors were conducted. In the paper the assessment of up-to-date experimental base status and it possibilities for further using in space nuclear energy field are given

  8. Experimental possibilities of research reactors complex Bajkal-1 for the decision of the problems of atomic power

    International Nuclear Information System (INIS)

    Research reactors complex 'Bajkal' includes two research reactors IVG.1M and RA. The reactor IVG.1M is a research water-water heterogeneous tank type nuclear reactor on the thermal neutrons with light-water moderator and coolant and beryllium neutron reflector. At present time the experimental studies of processes of the fission yield, the precipitation, the filtration of fission products. The possibilities of this reactor and stand systems are allowed to begin the experimental studies of model fuel assemblies water-cooled reactors at the accidental regimes. The reactor RA is a research high temperature gas-cooled tank type nuclear reactor on the thermal neutrons with gas moderator, zirconium hydride coolant and beryllium neutron reflector. The reactor RA is used for studies of hard-working of fuel elements and fuel assemblies of gas-cooled reactors during long reactor irradiation and experimental study of yield processes, precipitation and filtration of fission products

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Lee, C. S.; Seo, C. G

    2003-07-01

    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. 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

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. Irradiation of the experimental fuel assemblies with uranium-plutonium fuel in the BN-600 reactor

    International Nuclear Information System (INIS)

    Design features of experimental fuel assemblies (EFA) with uranium-plutonium mixed oxide fuel specific aspects of their arrangement within the BN-600 reactor core, conditions and basic results of EFA with the fuel mentioned in the BN-600 reactor are described

  15. An Ultrasonic Non Destructive Experimental System for the IRR2 Reactor Tank Wall

    International Nuclear Information System (INIS)

    The development and the validation of a new ultrasonic Non Destructive Experimental system in-service inspection of a research reactor closed tank are surveyed. In addition, preliminary results from the inspection of the IRR2 reactor tank are presented. The inspection has shown that the aluminium tank is in good physical condition with negligible corrosion and no significant structural defects

  16. Operating and test experience with Experimental Breeder Reactor number 2 (EBR-II), the Integral Fast Reactor (IFR) prototype

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor number 2 (EBR-II) has operated for 30 years, the longest for any liquid metal cooled reactor (LMR) power plant in the world. Given the scope of what has been developed and demonstrated over those years, it is arguably the most successful test reactor operation ever. Tests have been carried out on virtually every fast reactor fuel type. The reactor itself has been extensively studied. The most dramatic safety tests, conducted on 3 April, 1986, showed that an LMR with metallic fuel could safely accommodate loss of flow or loss of heat-sink without scram. EBR-II operated as the Integral Fast Reactor (IFR) prototype, demonstrating important innovations in safety, plant design, fuel design and actinide recycle. The ability to accommodate anticipated transients without scram passively resulted in significant simplification of the reactor plant, primarily through less reliance on emergency power and not having to require the secondary sodium or steam systems to be safety grade. These features have been quantified in a probabilistic risk assessment (PRA) conducted for EBR-II, demonstrating considerable safety advantages over other reactor concepts. Fundamental to the superior safety and operating characteristics of this reactor is the metallic U-Pu-Zr alloy fuel. Performance of the fuel has been fully proven: achieved burnup levels exceed 20 at.% in the lead test assemblies. A complete set of fuel performance and safety limits has been developed and was carried forward in formal safety documents supporting conversion of the core to IFR fuel. The last major demonstration planned was to assess the performance of recycled actinides in the fuel and to confirm that passive safety characteristics are maintained with recycled actinide fuel in the core. (author)

  17. 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

  18. 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.

  19. Concept, experimental and calculational investigations of a micromodule reactor

    International Nuclear Information System (INIS)

    In addition to the evolutionary improvements and development of new additional devices safety systems, much prominence is now given to the search of technical principals providing the enhanced safety for a NPP using new reactor design concepts as a whole. One of such concepts refers to the experience of the Institute of Physics and Power Engineering (IPPE), accumulated in the area of the exploitation of reactor channel loops for research purposes. A long-term experience of using such loops reveals that they feature a high reliability, safety and high technical and engineering parameters. (author)

  20. 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)

  1. 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

  2. Conceptual design study of quasi-steady state fusion experimental reactor (FEQ-Q), part 1

    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. Starting from 1984 JER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included; core plasma, reactor structure, reactor core components, magnets. (author)

  3. Analysis of the experimental neutron noise from the PHENIX reactor

    International Nuclear Information System (INIS)

    This paper deals with the interpretation of the neutron noise measurements in sodium-cooled fast neutron reactors and the problem of the core vibration monitoring. Given that the compaction of a SFR core results in a positive reactivity coefficient, monitoring any core movement is mandatory. Although several sodium-cooled reactors have been operated throughout the world, only the French PHENIX reactor matched the needs in terms of instrumentation and available data. This paper presents an analysis oriented towards the core compaction monitoring, of the measurements performed on the PHENIX reactor recorded in the SARA system. The main result is the observation of the neutron noise spectra as a function of power: as already proposed in the early years of PHENIX and SUPERPHENIX (SPX), the spectrum reflects the mechanical vibration of the fuel assembly lattice. The cross correlation with measurements such as vibration, sonar and temperature do not provide significant additional information to confirm or disprove this interpretation of the neutron noise spectrum: the temperature fluctuations lie in a distinct frequency domain, the sonar and vibration measurements on the control rods suffer from high detection noises. This paper also highlights that the interpretation of the noise measurements depends on the recording of the raw data, allowing data post analysis, development of new interpretative techniques, and feedback in terms of design of the instrumentation. As such, the SARA system is an example to reproduce, as far as the sustainability of the knowledge is concerned. (authors)

  4. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    In response to the accident of TEPCO Fukushima Daiichi Nuclear Power Station, the government took actions through the revision of regulatory standards as well as the complete separation of regulation administrative department from promotion administrative department. The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry, which has been in charge of the regulations of commercial reactors, and the Office of Nuclear Regulations of the Ministry of Education, Culture, Sports, Science and Technology, which has been in charge of the regulations of reactors for experiment and research, were separated from both ministries, and integrated into the Nuclear Regulation Authority, which was newly established as the affiliated agency of the Ministry of the Environment. As for the revision of regulations and standards, the Nuclear Safety Commission was dismantled, and regulation enacting authority was given to the new Nuclear Regulation Authority, and the regulations that stipulated new regulatory standards were enacted. This paper outlines the contents of regulations related mainly to the reactors for experiment and research, and explains the following: (1) retroactive application of the new regulatory standards to existing reactor facilities, (2) examinations at the Nuclear Regulatory Agency, (3) procedures to confirm the compliance to the new standards, (4) seismic design classification, and (5) importance classification of safety function. (A.O.)

  5. Some particular problems put by operating experimental reactors

    International Nuclear Information System (INIS)

    On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author)

  6. 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

  7. A proposal of reactor physics research of accelerator drive system using transmutation physics experimental Facility

    International Nuclear Information System (INIS)

    Reactor physics section of the Atomic Energy Society of Japan (AESJ) recognizes an accelerator driven system (ADS) as the next generation reactor and to promote researches using it. History of this section activity on ADS, outline of Transmutation Physics Experimental Facility in the 'High-Intensity Proton Accelerator Project', a proposal of reactor physics section to the project and future actions of this section are explained. The Transmutation Physics Experimental Facility consists of a fast neutron subcritical system and a nuclear spallation neutron source. The contents of experiments are evaluation of nuclear properties of fast neutron subcritical system driven by nuclear spallation source, verification of operation and control of accelerator driven hybrid system and evaluation of nuclear transmutation characteristics of MA (Minor Actinides) and LLFP (Long-Lived Fission Product). Themes of R and D of ADS contain operation control of ADS, critical control of subcritical system, properties of reactor with nuclear spallation neutron source and nuclear transmutation characteristics. The experimental items are measurement of dynamic characteristics of reactor at beam change, R and D of method of output control and stop, R and D of contentious monitoring method of subcritical multiplication, measurement of dynamic characteristics of behaviors of reactivity, effects on reactor characteristics of high energy neutron, effects on reactor physics of beam duct and large target, nuclear transmutation efficiency and simulation of nuclear transmutation reactor core. (S.Y.)

  8. 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)

  9. 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

  10. 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.

  11. 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

  12. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 9500C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

  13. Experimental study of fluidic mixing in a cylindrical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Orfaniotis, A.; Fonade, C.; Lalane, M.; Doubrovine, N. [Centre National de la Recherche Scientifique (CNRS), 31 - Toulouse (France)

    1996-04-01

    Fluidic mixing in a cylindrical reactor was studied in an effort to determine the effect of jet disposition and the viscosity of the liquid. The tests were carried out in a a tank using conductimetric probes to measure the mixing time. Results indicated that relative jet positions leading to an impinging flow structure were less efficient than shear flow configurations. When these results were compared with results of earlier work by Simon and Fonade (1993) it was found that they were consistent with the exponent 2/3 obtained by them in experiments with turbulent jets. It was pointed out that these mixing times apply only to mixing in cylindrical reactors. With different geometries, such as basins and lagoons with small liquid depths, a new choice of the reference length included in the expression of the reference time will be needed. 10 refs., 3 tabs., 16 figs.

  14. Design project of the experimental facility for testing uranium creep in the reactor

    International Nuclear Information System (INIS)

    This report contains the design for constructing the experimental device for testing metal uranium creep in the RA reactor core under defined neutron flux conditions, temperature, mechanical loads and time of irradiation. This device will be placed in one of the experimental channels in the core. This report contains physical, thermal and mechanical calculations and engineering drawings of the device

  15. 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

  16. 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.

  17. 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.

  18. 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)

  19. Simulation in an experimental reactor of a load follow and a remote control operation

    International Nuclear Information System (INIS)

    In order to confirm fuel reliability for operation in daily load follow and remote control, a simulation of this kind of operation was carried out in a CEA experimental reactor containing twelve 17x17 assemblies. The manner with which this simulation has been defined on the basis of operation data on a power reactor and its implementation is presented. Load variations corresponded to a cycle consisting of a decrease in ower, a plateau at reduced power and a power escalation to full rated power. The definition of the irradiation conditions in the experimental reactor was based on the same power requirements as those obtained in a 900 MWe reactor operating under load follow and remote control

  20. Project management and its characteristic analysis for China experimental fast reactor

    International Nuclear Information System (INIS)

    As the first step of fast reactor development in China, China Experimental Fast Reactor (CEFR), one of the key projects of the National High Technology Research and Development Program (863 Program), will reach the first physical criticality in 2009 and will be connected to grid in 2010. The CEFR project includes R and D, design, construction, commissioning and operation, and its technology and management are very complex. This paper describes its position, objective, task, management, and analyzes its characteristics. (authors)

  1. Linear and nonlinear stability analysis, associated to experimental fast reactors. Part 2

    International Nuclear Information System (INIS)

    The nonlinear effects in fast reactors kinetics and their stability are studied. The Lyapunov criteria and the Lurie-Letov functions for nonlinear systems were established and simulated. Small oscillations were studied by a Fourier analysis to clarify particular aspects of feedback and load functions in fast reactor at zero power, or/and in normal power level. The results were in agreement with the experimental data existing in the literature. (E.G.)

  2. Survey of experimental studies on release and deposition of reactor fuel fission products

    International Nuclear Information System (INIS)

    In the work, a review of the most important results from 11 series of experimental studies on fission product behaviour in reactor accident conditions that were performed in a number of research facilities worldwide, is presented. The facilities can be divided into out-of-pile, in-pile and integral ones, the later ones modelling the whole of a reactor cooling system. Emphasis is given not only on quantitative description of release and deposition phenomena but on physico-chemical processes accompanying radioactivity migration in reactor circuits and variety of FP chemical forms as well. (author)

  3. Measurements of gamma-ray energy deposition in a heterogeneous reactor experimental configuration and their analysis

    International Nuclear Information System (INIS)

    An important contribution to the power output of a fast reactor is provided by the energy deposition from gamma-rays, and is particularly significant in the inner fertile zones of heterogeneous breeder reactor designs. To establish the validity of calculational methods and data for such systems an extensive series of measurements was performed in the zero power reactor Masurca, as part of the RACINE programme. The experimental study involved four European laboratories and the measurement techniques covered a range of thermoluminescent dosemeters and an ionization chamber. The present paper describes and compares the gamma-ray energy deposition measurements and analysis

  4. Neutronics and activation analysis of a SS316 based experimental D-T fusion power reactor

    International Nuclear Information System (INIS)

    Neutronics and activation analysis for a SS316 based experimental tokamak fusion power reactor was completed using a two-dimensional neutronics model because of its practical features. Operating and decay power densities in all reactor components were obtained. The significant increase of neutron fluxes in the vacuum duct region was revealed. Maintenance and waste management aspects of the reactor components were investigated. Impacts on these aspects due to important minor alloying elements and impurities such as Mn, Co, Mo, and Nb were addressed

  5. Proposal for an international experimental pebble bed reactor - HTR2008-58174

    International Nuclear Information System (INIS)

    HTRs, both prismatic block fuelled and pebble fuelled, feature a number of uniquely beneficial characteristics that will be discussed in this paper. In this paper the construction of an international experimental pebble bed reactor is proposed, possible experiments suggested and an invitation extended to interested partners for co-operation in the project. Experimental verification by nuclear regulators in order to facilitate licensing and the development of a new generation of reactors create a strong need for such a reactor. Suggested experiments include: Optimized incineration of waste Pu in a pebble bed reactor: The capability to incineration pure reactor grade plutonium by means of ultra high burn-up in pebble bed reactors will be presented at this conference in the track on fuel and fuel cycles. This will enable incineration of the global stockpile of separated reactor grade Pu within a relatively short time span. Testing of fuel sphere geometries, aimed at improving neutron moderation and a decrease in fuel temperatures. Th/Pu fuel cycles: Previous HTR programs demonstrated the viability of a Th-232 fuel-cycle, using highly enriched uranium (HEU) as driver material. However, considerations favoring proliferation resistance limit the enrichment level of uranium in commercial reactors to 20 %, thereby lowering the isotopic efficiency. Therefore, Pu driver material should be developed to replace the HEU component. Instead of deploying a (Th, Pu)O fuel concept, the proposal is to use the unique capability offered by pebble bed reactors in deploying separate Th- and Pu-containing pebbles, which can be cycled differently. Testing of carbon-fiber-carbon (CFC) structures for in-core or near-core applications, such as guide tubes for reserve shutdown systems, thus creating the possibility to safely shutdown reactors with increased diameter. Development of very high temperature reactor components for process heat applications. Advanced decay heat removal systems e

  6. Experimental analysis on thermohydraulic characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction mass flow rate and stability of the system, especially at lower pressure; secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit does not reach its situation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0K<ΔT<28 K) there exists three regions for system flow condition, namely, stable two-phase flow, bulk and subcooled boiling unstable flow and subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, are shown in the above inlet subcooling range, and based on it the instability map of the system are given through experiment and calculation

  7. 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)

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

    International Nuclear Information System (INIS)

    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 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-1, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

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

    International Nuclear Information System (INIS)

    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 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-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)

  10. Automatic control of neutron flux in experimental channels of the WWR-M type reactors

    International Nuclear Information System (INIS)

    The flowsheet of the neutron flux local regulator intended for maintaining the given level of neutron flux distribution in experimental channels of the WWR-M type reactor under stationary and transition modes, is suggested. The functional diagram of the electron regulation block (ERB) in considered. The regulator is tested when the reactor operates with the capacity of 13 MWt along with the staff system of automated regulation and without it. The experiments carried out demonstrate the stable operation of the entire control system and good performance characteristics of the ERR block. The conclusion is made that the suggested method of neutron flux automated regulation in experimental channels can be successfully extended to a higher number of experimental channels and applied at other research reactors. Small size fission chambers and direct charging detectors can be used in local systems as sensors

  11. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    International Nuclear Information System (INIS)

    Highlights: • Custom-built high temperature, high pressure tribometer is designed. • Two different wear phenomena at high temperatures are observed. • Experimental wear results for graphite are presented. • The graphite wear dust production in a typical Pebble Bed Reactor is predicted. -- Abstract: This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor

  12. 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.)

  13. 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)

  14. The inverse kinetics technique for reactor shutdown measurement -an experimental assessment

    International Nuclear Information System (INIS)

    It is proposed to use the Inverse Kinetics Technique to measure the subcritical reactivity as a function of time during the testing of the nitrogen injection systems on AGRs. An experimental assessment of the technique by investigating known transients created by control rod movements on a small (2m high, 1m radius) experimental reactor is described. Spatial effects were observed close to the moving rods but otherwise derived reactivities were independent of detector position and agreed well with the existing calibrations. This prompted the suggestion that data from installed reactor instrumentation could be used to calibrate CAGR control rods. (author)

  15. 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

  16. Experimental studies on sodium fires. Application to reactors

    International Nuclear Information System (INIS)

    The experimental program developed at the Cadarache Center by the Department of Nuclear Safety is described. The most important part of the program relates to great pool fires occuring in tight or poorly ventilated vessels. The purpose is to simulate sodium leakage resulting from pipe rupture in the secondary cooling system. The program has two parts. Cassandre for pool fires at 550 deg C and Lucifer for pool fires of 300 kg over 2m2

  17. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    In 1969 JAERI started the design study of the Experimental Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR), and trial design, preliminary design, conceptual design, comprehensive system design and the first and second stage of detailed design have been carried out. Hereafter JAERI is going to pursue the rationalized Experimental VHTR system which maintains the required functions and performance and has the potential for reducing the construction cost, utilizing extensively the inherent safety features of HTGRs. In the current design, i.e. the second stage of detailed design, the reactor outlet coolant temperature is 9500C to aim earlier construction of the Experimental VHTR, according to the specification in ''Long-term plan for the development and utilization of nuclear energy'' revised by Japan Atomic Energy Commission in June 1982. This report presents the results based mainly on the comprehensive system design (completed by 1980.3) which is the last overall system design of the Experimental VHTR aiming 10000C reactor outlet coolant temperature and partially on the first stage (completed by 1981.3) of detailed design in the form of ''an application of reactor construction permit, Appendix 8'', excepting comformance with ''Safety Design Requirements'' which correspond to ''Safety Design Criteria for Water Cooled Nuclear Power Plants issued by Japan Nuclear Safety Commission''. (author)

  18. 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.

  19. 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

  20. 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.

  1. 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

  2. Experimental RA reactor operation with 80% enriched fuel - Program of experimental operation: a) Program of experimental operation with 80% enriched fuel at low power, b) contents of the experimental operation with 80% enriched fuel at higher power levels

    International Nuclear Information System (INIS)

    Highly enriched (80%) uranium oxide fuel was regularly used in the mixed reactor core with the 2% enriched fuel since 1976. The most important changes related to reactor operation, in comparison with the original design project were related to reactor core fuelling schemes. At the end of 1979 reactor was shutdown due to the corrosion coating noticed on some fuel elements and due to decrease quality of the heavy water. Subsequently the Sanitary inspector of Serbia has prohibited further reactor operation. Restart of the reactor will not be a simple continuation of operation. It is indispensable to perform complete experimental program including measurements of critical parameters at different power levels for the core with fresh 80% enriched fuel. The aim of this document is to obtain working permission and its contents are in agreement with the procedure demanded by the Safety Committee of the Institute. It includes results of optimization and safety analysis for the initial reactor core. Since the permission for restart is not obtained, a separate RA reactor safety report is prepared in addition to the program for experimental operation. This report includes: detailed program for reactor experimental operation with 80% enriched fuel in the core at low power levels, and contents of the experimental operation with 80% enriched fuel in the core at higher power levels

  3. 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)

  4. Experimental evaluation of the seismic capacity of VVER 440-213 type reactor control rod system

    International Nuclear Information System (INIS)

    The experimental evaluation of the WWER-440/213 control rod drive seismic capacity was carried out on the CKTI-VIBROSEISM Horizontal Shaking Table, specifically designed for seismic testing of the full-scale control rod drive of WWER-440 and WWER-1000 reactors. A detailed description is given of the experimental conditions and the methodology used. The results are compared with technical demands on control rods in emergency situations. (Z.S.) 4 refs

  5. Experimental software design of neutron texture diffractometer at China advanced research reactor

    International Nuclear Information System (INIS)

    The experimental software of the neutron texture diffractometer at China Advanced Research Reactor (CARR) was designed. Based on the principle of texture measurement by neutron diffraction and the motion control and data acquisition system of the diffractometer, the functions needed for texture measurement were proposed. Then the flow charts of these functions were described in detail and realized by Python language in Linux system. The experimental software for CARR neutron texture diffractometer has been successfully accomplished. (authors)

  6. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Goncharov, V.V.; Dubrovin, K.P.; Ivanov, E.G.; Korneev, V.T.; Kruglov, A.B.; Lebedev, L.M.

    1987-11-01

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < ..mu..m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).

  7. Tests in an MR reactor of experimental fuel elements of the VVER-1000 reactor

    International Nuclear Information System (INIS)

    In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVER-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory. The test in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated the fuel elements of the specified design (with initial helium pressures of 1.96-2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film < μm thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level)

  8. 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)

  9. Measurement of spectrum at the experimental 6.5 MW reactor in Vinca

    International Nuclear Information System (INIS)

    Since RA reactor is supplied with horizontal experimental channels which lead directly to the core fast neutron spectrum in the channel does not differ much from the neutron spectrum in the core. Spectrum was measured by 'telescope' for detecting scattered protons. Measuring procedure together with the measured spectrum are presented in this paper

  10. International Thermonuclear Experimental Reactor (ITER). Engineering Design Activities (EDA). Agreement and protocol 1

    International Nuclear Information System (INIS)

    This document contains protocol 1 to the agreement among the European Atomic Energy Community, the government of Japan, the Government of the Russian Federation, and the Government of the United States of America on cooperation in the engineering design activities for the International Thermonuclear Experimental Reactor, which activities shall be conducted under the auspices of the International Atomic Energy Agency

  11. Out-of-pile experimental stand for research of fast reactors safety questions

    International Nuclear Information System (INIS)

    In the given paper there is given the description of experimental facility 'EAGLE', which is meant for research of safety problems of fast reactors one of which is excluding of re-criticality in case of severe accident with core melting. There are demonstrated concepts and volume of planned tests and also the results of conducted experiments. (author)

  12. 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...

  13. 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...

  14. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II

    International Nuclear Information System (INIS)

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material, that is, more than ''invested,'' and at the same time produce power. The group identified the problem of

  15. Civilian nuclear power on the drawing board: the development of Experimental Breeder Reactor-II.

    Energy Technology Data Exchange (ETDEWEB)

    Westfall, C.

    2003-02-20

    On September 28, 2001 a symposium was held at Argonne National Laboratory as part of the festivities to mark the 100th birthday of Enrico Fermi. The symposium celebrated Fermi's ''contribution to the development of nuclear power'' and focused on one particular ''line of development'' resulting from Fermi's interest in power reactors: Argonne's fast reactor program. Symposium participants made many references to the ways in which the program was linked to Fermi, who led the team which created the world's first self-sustaining nuclear chain reaction. For example, one presentation featured an April, 1944 memo that described a meeting attended by Fermi and others. The memo came from the time when research on plutonium and the nuclear chain reaction at Chicago's WWII Metallurgical Laboratory was nearing its end. Even as other parts of the Manhattan Engineering Project were building on this effort to create the bombs that would end the war, Fermi and his colleagues were taking the first steps to plan the use of nuclear energy in the postwar era. After noting that Fermi ''viewed the use of [nuclear] power for the heating of cities with sympathy,'' the group outlined several power reactor designs. In the course of discussion, Fermi and his colleagues took the first steps in conjuring the vision that would later be brought to life with Experimental Breeder Reactor I (EBR-I) and Experimental Breeder Reactor II (EBR-II), the celebrated achievements of the Argonne fast reactor program. Group members considered various schemes for a breeder reactor in which the relatively abundant U-238 would be placed near a core of fissionable material. The reactor would be a fast reactor; that is, neutrons would not be moderated, as were most wartime reactors. Thus, the large number of neutrons emitted in fast neutron fission would hit the U-238 and create ''extra'' fissionable material

  16. Safety design implementation for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    A high level of safety is being integrated into the ITER design. This paper describes some of the steps being followed in the facility, system and component design to ensure safety. The safety approach developed for ITER takes into account the moderate hazard associated with ITER, its role as an experimental facility, and draws on experience in nuclear and non-nuclear industries. A key element is a graded approach to match requirements to hazards being controlled. This is reflected in, for example, dose limits that are lower for higher frequencies, classification of components in terms of their importance to safety in order to guide the setting of requirements, and structural design. The classification system that ITER is developing for safety-relevant components is described. the implementation of this classification into the design is still very much under development, but preliminary thoughts are outlined here. Processes are in place to determine the safety functions needed to ensure public safety, to identify systems that fulfill these safety functions, to set system requirements to ensure these functions are implemented in the design, to design system components to meet requirements, and to identify design, manufacture and operations requirements needed. At this stage of the project, it is concluded that the design and operation could meet the safety-related requirements of any of the potential host countries with only minimal modifications to accommodate the characteristics of the specific site chosen. (author)

  17. Report on the operation of the FMRB reactor and related activities in the year 1981

    International Nuclear Information System (INIS)

    In 1981 the Forschungs- und Messreaktor Braunschweig (FMRB) had been in action for 2400 hours without any serious disturbance. The released power for this period amounts to 2294 MWh. The experience in the field of reactor operation, in the radiation protection work as well as in the experiments at the beam tubes is reported. (orig.)

  18. 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.

  19. Current state of peripheral equipment of reactor. Installation of new experimental apparatus and decommissioning measures

    International Nuclear Information System (INIS)

    Kyoto University Research Reactor Institute, which celebrated its 50th anniversary in 2013, holds many obsolete reactor peripheral devices that were also installed at the time of reactor installation, which has posed obstacles to many studies. Therefore, toward the utilization efficiency improvement and new research creation, the university is promoting the reform of the reactor peripheral devices. On the other hand, in response to the accident of TEPCO Fukushima Daiichi Nuclear Power Station, the safety of reactor was sought more severely, which obliged the decommissioning measures of old experimental facilities. This paper reports the decommissioning measures of cold neutron source facilities, with a focus on the overview of the equipment, processing methods for deuterium gas under use, demonstration experiment for used hydrogen gas processing unit, and processing results. It takes up the installation of positron beam apparatus, special neutron lab, and small multi-purpose neutron diffractometer, as new experimental facilities, and introduces the outline of the devices, utilization purpose, and features of the devices. (A.O.)

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

    International Nuclear Information System (INIS)

    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.

  1. Current status of experimental breeder reactor-II [EBR-II] shutdown planning

    International Nuclear Information System (INIS)

    The Experimental Breeder Reactor--II (EBR-II) at Argonne National Laboratory--West (ANL-W) in Idaho, was shutdown in September, 1994 as mandated by the US Department of Energy. This sodium cooled reactor had been in service since 1964, and was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the Sodium Process Facility. The sodium environment and the EBR-II configuration, combined with the radiation and contamination associated with thirty years of reactor operation, posed problems specific to liquid metal reactor deactivation. The methods being developed and implemented at EBR-II can be applied to other similar situations in the US and abroad

  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. Development of a numerical tool for safety assessment and emergency management of experimental reactors

    International Nuclear Information System (INIS)

    The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. Among its duties, one important item is to provide help for emergency situations management in case of an accident occurring in a French nuclear facility. In this framework, IRSN develops and applies numerical tools dealing with containment management issues. Up to now IRSN has not got any specific tool for experimental reactors. Accordingly, it has been then decided to extend the ASTEC code, devoted to severe accident scenarios for Pressurized Water Reactors, to this kind of reactors. This lumped-parameter code, co-developed by IRSN and GRS (Germany), covers the entire phenomenology from the initiating event up to fission products release outside the reactor containment, except for the steam explosion and the mechanical integrity of the containment. A first application to experimental reactors was carried out to assess the High Flux Reactor (HFR) operator's improvement proposal concerning the containment management during accidental situations. This reactor, located in Grenoble (France), is composed of a double wall containment with a pressurized containment annulus preventing any direct leakage into the environment. Until now, in case of severe accidents (mainly core melting in pool, explosive reactivity accident called BORAX), the HFR emergency management consisted in isolating the containment building in the early stage of the accident, to prevent any radioactive products release to the environment. The operator decided to improve this containment management during accidental situations by using an air filtering venting system able to maintain a slight sub-atmospheric pressure in the reactor building. The operator's demonstration of the efficiency of this new system is mainly based on containment pressure evaluations during accidental transients. IRSN assessed these calculations through ASTEC calculations. Finally, a global agreement was

  5. 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)

  6. 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)

  7. 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)

  8. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

    International Nuclear Information System (INIS)

    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.)

  9. 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.)

  10. 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

  11. 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)

  12. Molten salt reactors. Synthesis of studies realized between 1973 and 1983. Experimental loop file

    International Nuclear Information System (INIS)

    Four test loops were developed for the experimental study of a molten salt reactor with lead salt direct contact. A molten salt loop, completely in graphite, including the pump, showed that this material is convenient for salt containment and circulation. Reactor components like flowmeters, electromagnetic pumps, pressure gauge, valves developed for liquid sodium, were tested with liquid lead. A water-mercury loop was built for lead-molten salt simulation studies. Finally a lead-salt loop (COMPARSE) was built to study the behaviour of salt particles carried by lead in the heat exchanger

  13. Measurements of gamma-ray energy deposition in a heterogeneous reactor experimental configuration and their analysis

    International Nuclear Information System (INIS)

    The in-core gamma-ray energy deposition contributes significantly to the power output of a fast reactor. The designer of a heterogeneous fast breeder reactor needs a reliable calculation system to evaluate the gamma heating, in particular in the inner breeder zones, where it represents a large fraction of the power. In order to check the calculation system, measurements were made with various experimental techniques in a RACINE assembly of the MASURCA critical facility, simulating a large double-annulus heterogeneous FBR

  14. Experimental investigations of actinide release from coated fuel particles for high-temperature reactors

    International Nuclear Information System (INIS)

    The migrational behaviour of actinides in the coated fuel particles proposed for high-temperature reactors is investigated experimentally. Data are described in the framework of the diffusion model. The experimental procedures are presented and the necessary computer codes are discussed. The diffusion coefficients of the actinides - plutonium, americium and curium - as well as of the fission product cesium are derived from the experimental data by a nonlinear least squares fit procedure and are presented in the form of Arrhenius lines D = Do esup(-Q/RT) for U(Th)-O2, HTI-PyC and SiC. (orig.)

  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. Extracted neutron beams experimental facilities and program of the first experiments at the IBR-2 reactor

    International Nuclear Information System (INIS)

    Structural specific features of the IBR-2 pulse research biological hidered reactor. The characteristics of spectrometer for investigating the small angle neutron scattering and the CORA facility intended for investigating the structure and dynamics of condensed media by means of the therrol neutron scattering as well as the DN-2 diffractometer for investigating the atomic structure and crystallographic characteristics of monocrystals, having large (>10 A) elementary cell size and the equipment of the ultracold neutron channel are given. Biological shields of the reactor and experimental facilities are assembled of concrete blocks and standard building constructions and attains in the most dangerous regions 1 m. The shield ensures for the personnel a safety level of ionizing radiations and effectively shields the facilities from mutual effects caused by scattered radiation. The program of physical investigations planned at the IBR-2 continues the investigations started at the IBR-30 reactor

  17. 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

  18. 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.

  19. The D and D of the Experimental Boiling Water Reactor (EBWR)

    International Nuclear Information System (INIS)

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

  20. 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

  1. Experimental thermal-hydraulic analysis of the IPR-R1 TRIGA nuclear reactor

    International Nuclear Information System (INIS)

    The heat generated by nuclear fission in the IPR-R1 nuclear reactor is transferred from fuel elements to the cooling system through the fuel/cladding (gap) and the cladding to coolant interfaces. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally. A correlation for the gap conductance between the fuel and the cladding was also presented. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. Results indicated that subcooled boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 k W. (author)

  2. 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

  3. Project plan for the decontamination and decommissioning of the Argonne National Laboratory Experimental Boiling Water Reactor

    International Nuclear Information System (INIS)

    In 1956, the Experimental Boiling Water Reactor (EBWR) Facility was first operated at Argonne National Laboratory (ANL) as a test reactor to demonstrate the feasibility of operating an integrated power plant using a direct cycle boiling water reactor as a heat source. In 1967, ANL permanently shut down the EBWR and placed it in dry lay-up. This project plan presents the schedule and organization for the decontamination and decommissioning of the EBWR Facility which will allow it to be reused by other ANL scientific research programs. The project total estimated cost is $14.3M and is projected to generate 22,000 cubic feet of low-level radioactive waste which will be disposed of at an approved DOE burial ground. 18 figs., 3 tabs

  4. Analysis of Tokamak Fusion Test Reactor (TFTR) Prototype of International Thermonuclear Experimental Reactor (ITER)‡

    Science.gov (United States)

    Hester, Tim; Maglich, Bogdan; Scott, Dan; Calsec Collaboration

    2015-11-01

    TFTR produced world record of 10 million watts of controlled fusion power1 (CFP-1994) was summarized in Review1. We present evidence3 that: (1) TFTR input vs. output was 40 to 10 MW i.e. a power loss. (2) Review claims no experimental evidence for thermonuclear CFP production (only a calculation). (3) Ultra-high vacuum (UHV) required for τE = 0.2 s is 10-9 torr. TFTR had no UHV pumps, resulting in 10-3 torr, restricting τE fusion neutron power'' without particle energy identification, energy or coincidence. (6) 8 of 9 parameters claimed were inferred not measured. Quadratic test of TFTR data results2 in zero thermonuclear fusion power contribution to 10 MW: SFP = (0 +/- 1)%. ‡ Submitted to Physics of Plasmas†

  5. 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

  6. Testing of experimental fuel elements for VVER-1000 reactors in MR to high fuel burnup

    International Nuclear Information System (INIS)

    Pressurized water reactors are given a commanding role in the development program for the nation's nuclear power industry. Considerable operating experience has been gained with VVER-1000 reactors. As of the start of 1990, 17 units with VVER-1000 reactors were in operation in this country and abroad. The first loadings were designed for a 2-year run with average fuel burnup of 28.5 MW-day/kg. The rod-type fuel elements used in the reactors displayed high serviceability and reliability (leakage does not exceed 0.02%). Operating experience and the results of computational and experimental work have made it possible to substantiate the possibility of switching them to a 3-year run. The fifth unit of the Novovoronezh Atomic Power Plant was the first to be switched to a 3-year run, as of 1984. The average fuel burnup achieved after three fuel cycles was 42.6 MW-day/kg. All units with VVER-1000 reactors are now being switched to a 3-year run with an average burnup of more than 40 MW-day/kg for the unloaded fuel

  7. Experimental and computational studies of thermal mixing in next generation nuclear reactors

    Science.gov (United States)

    Landfried, Douglas Tyler

    The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

  8. 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)

  9. 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.

  10. Development and Optimization of Nuclear Heating Measurement Techniques in Zero Power Experimental Reactors

    International Nuclear Information System (INIS)

    The objective of this study is to develop nuclear heating measurement methods in Zero Power experimental reactors. This paper presents the analysis of Thermo- Luminescent Detectors (TLDs) and Optically Stimulated Luminescent Detectors (OSLDs) experiments in the MINERVE research reactors at the French Atomic Energy and Alternative Energies Commission center in Cadarache. The experimental sources of uncertainties on the dose have been reduced by using the optimum conditions of charged particle equilibrium (CPE) of the calibration step and reactor measurement for each detector types; by improving the process of the TLDs/OSLDs reading and calibration processes. The interpretation of these measurements needs to take into account several correction factors related to both the environment of calibration step and the type of detectors used. Similarly, the correction due to the neutrons contributions to the total dose integrated by the detectors is evaluated with Monte Carlo calculation methods. These calculations are based on MCNP simulations of neutron-gamma and gamma-electron transport coupled particles using ENDF/B-VI nuclear data library. TLDs and OSLDs are positioned inside aluminum or hafnium pillboxes. Comparisons between calculated and measured integral gamma-ray absorbed doses by TLDs in these new experiments are carried out in the MINERVE reactor in the surrounding aluminum material. They show that calculations slightly overestimate the measurements by about 8 %. By using OSLDs, the calculation slightly overestimates the measurement by about 6 %. (authors)

  11. 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)

  12. LBE-water interaction in sub-critical reactors: First experimental and modelling results

    International Nuclear Information System (INIS)

    This paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 deg. C in a reaction vessel containing LBE at 1 bar and 350 deg. C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors

  13. Experimental and MCNP calculations of neutron flux parameters in irradiation channel at Es-Salam reactor

    International Nuclear Information System (INIS)

    The Algerian research reactor (Es-Salam) is a 15 MW heavy water reactor type, operating since 1992. It became essential to characterize the neutron field in the most useful irradiation positions, in order to guarantee the accuracy in the application of k0-neutron activation analysis (k0-NAA). Experimental value of the thermal to epithermal neutron flux ratio (f) and of the deviation of the epithermal neutron spectrum from 1/E shape (α) were determined using different methods. This work focuses the verification of Monte Carlo neutron flux calculation in typical irradiation channel. Comparison of the results for parameter f obtained experimentally and by Monte Carlo simulations shows good agreement in the irradiation channel studied. The difference between both results is about 2.08%. (author)

  14. 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).

  15. Driving requirements for the OHMIC heating and equilibrium coils of a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    The tokamak experimental power reactor will have substantial power driving requirements, on the order of 1000 MW for the ohmic heating system, and 300 MW for the equilibrium system. A combined energy of some 450 MJ must be supplied to start the burn cycle, and a flux change of roughly 35 V-sec is required at both coils. Different plasma or operating conditions can increase these requirements, sometimes substantially

  16. Design study of superconducting poloidal magnets for a tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design of superconducting poloidal magnets for a tokamak experimental fusion reactor has been studied. The purpose is to reveal engineering problems of poloidal magnets. Electrical insulation at liquid helium temperature, fatigue of reinforcement materials of the windings due to repeated electromagnetic forces, construction of FRP liquid helium containers and transfer loss of liquid helium are found to be especially important. The schedule of R and D is also presented. (auth.)

  17. 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)

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

    OpenAIRE

    Agnieszka Kaczmarek-Kacprzak; Marcin Jaskólski

    2012-01-01

    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 ana...

  19. Analysis on vibration characteristics of the primary sodium pump of China experimental fast reactor

    International Nuclear Information System (INIS)

    The article introduces the rotational model analysis and the vibration test of the primary sodium pump of China Experimental Fast Reactor. Through the establishment of the rotation of the shaft system model, the critical speed has been analyzed. Combined with the pump bearings and motor bearings at double amplitude and RMS vibration velocity measurement experiment, the results show that the vibration characteristics of the primary sodium pump meet the requirements of the operational limits. (author)

  20. Safety analysis of superconducting toroidal field magnet for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Safety analysis of the superconducting toroidal field magnet for a Tokamak experimental fusion reactor has been carried out. Works were accident classification, FMEA and FTA analyses, coil stability and quench behavior calculations, failure detection and coil protection system designs, structure analysis, fracture and fatigue studies, and earthquake response analysis. Accident analysis of cryostat and refrigeration system was also performed. The objective of this work is to reveal technological problems of the toroidal field magnet by safety analysis. (author)

  1. 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

  2. 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

  3. 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)

  4. 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.

  5. Transient simulation code development of primary coolant system of Chinese Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: ► A transient analysis code is developed for Chinese Experimental Fast Reactor. ► A set of subroutines for friction and heat transfer correlations were compiled. ► The calculation speed of this code is fast enough for real-time simulation. - Abstract: Chinese Experimental Fast Reactor (CEFR) is a 25 MWe sodium cooled, pool type reactor, which was built at the China Institute of Atomic Energy in Beijing as the forerunner to the first-stage of Chinese fast reactor development plans. In order to understand the response of the Primary Coolant System (PCS) to various transients and train the operators a dynamic model using basic energy and momentum equations was developed with some assumptions. Heat transfer models for reactor core and intermediate heat exchanger were also included. Subroutines were developed to calculate the thermal properties, friction coefficients and heat transfer coefficients of liquid sodium. Gear’s method was applied to solve the dynamic model. A transient analysis code named THPCS (Thermal–Hydraulic code of PCS) was developed and is independent of the design and safety analysis codes. Three typical events, such as loss of one primary pump, unprotected transient overpower and accidental closure of primary pump check valves were chosen and investigated. The prediction results of the code agree well with those of the final safety analysis report of CEFR. A fourth postulated accident, station blackout without scram and loss of all heat sink, was also analyzed to show the ability of the code, which is more serious than the former. The transient simulation code developed in this paper will be useful for the safety operation of CEFR

  6. Status and possibility of fuel and structural materials experimental irradiation in BN-600 reactor. Stages of BN-600 reactor core development

    International Nuclear Information System (INIS)

    The results of the irradiation of standard and experimental fuel subassemblies (SA) in BN-600 reactor are presented. The prospects of further tests on experimental SAs and on standard SAs up to 12% h.a. burnup and damage doses ≥ 90 dpa are also analyzed. (author)

  7. Experimental and modeling analysis of a batch gasification/pyrolysis reactor

    International Nuclear Information System (INIS)

    This paper presents some experimental results about biomass gasification obtained with a bench-scale gasifier consisting of an indirectly heated batch reactor, inserted in a high temperature furnace. Spruce wood has been used as feedstock in different forms and sizes (as pellets and sawdust). The experimental activity includes the analysis of the thermal response of the system (using an inert bed material) and the characterization of the gasification products. The yield of the gaseous compounds found in the syngas has been measured using an on-line gas-chromatography technique. The experimental results have been compared against calculations obtained by applying a thermochemical equilibrium model, improved to predict both the gas and the solid phase product yields. The experimental (average) yield of reaction products has been found to be in a satisfying agreement with the thermodynamic model.

  8. Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation)

    International Nuclear Information System (INIS)

    According to the demands for obtaining the licence for restarting the Ra reactor and the experimental operation this document includes the radiation monitoring measured data in the working space and environment of the RA reactor, i.e. Boris Kidric Institute. The meteorology measured data are included as well. All the measurements are performed according to the radiation protection program applied actually from the first reactor start-up at the end of 1959

  9. 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

  10. Research nuclear reactor and particle accelerator as complementary facilities in obtaining experimental nuclear data

    International Nuclear Information System (INIS)

    In the last decade a large amount of diverse and high precision nuclear data is in high demand to support both power applications (in nuclear fusion and fission reactors, fuel cycle in all its stages, nuclear safety) and non-power applications (radiation dosimetry, life sciences, ecology, industry, etc.) of atomic and nuclear techniques. The atomic and nuclear data are generated from experimental measurements, theoretical model calculations and data evaluation, which are finally validated internationally and included in data bases under standardized formats. Measuring of these data imply utilization of research reactors and charged particle accelerators, in complex experiments characterized by high degree of complementarity. Aspects of this complementarity in the nuclear data obtained from reactors and accelerators will be presented in this work. In Romania an advanced research reactor (TRIGA at INR Pitesti) and an electrostatic 4-5 MeV/nucleon accelerator (TANDEM Van de Graaff at IFIN - HH, Bucharest) are operational and a rich scientific expertise in the field of nuclear structure and reaction mechanisms is available. Consequently, the paper considers a project at a national scale for measuring and evaluating nuclear data. Having in view numerous signals launched by international organizations (IAEA-Vienna, NEA-OECD, NNDC-USA) such a project would have a powerful international support because of increasing world wide demand of atomic and nuclear data. Nuclear data are either structure and decay nuclear data or reaction nuclear data. The first class refers to nuclear state properties (masses, excitation energies, quantum numbers, lifetimes, etc.) as well as to their decay modes. Data from the second class refer to differential or integral cross sections. The paper presents comparatively the data obtainable at accelerators and reactors for the two above mentioned classes of nuclear data, particularly, the data required for building ADS (Accelerator Driven Systems

  11. Training and retraining of personnel working at experimental reactors in compliance with quality assurance obligations

    International Nuclear Information System (INIS)

    At the Saclay site near Paris the Commissariat a l'Energie Atomique is operating three experimental reactors - OSIRIS (70 MW), ORPHEE (14 MW) and ISIS (700 kW). These reactors are run by operators and maintenance personnel under the direction of specialized engineers. Operation is regulated by an official text which specifies that any facility head must have available the resources to ensure training of future staff before they are authorized to take up their posts, having first undergone a test of their behaviour and knowledge. The Saclay Reactor Service instituted procedures four years ago which defined the rules concerning recruitment, training, retraining, and authorization of various staff members, placing particular emphasis on permanent staff. Training and retraining of personnel is done in collaboration with the National Institute for Nuclear Science and Technology at Saclay, which awards an official diploma for a course on nuclear reactor control and instrumentation, with the Grenoble Technological Simulation and Training Centre, and with the Training Service of the Saclay Nuclear Centre

  12. 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

  13. 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...

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

    International Nuclear Information System (INIS)

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

  15. 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.

  16. 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.

  17. Design, Fabrication and Testing of the Control Rods for the Experimental Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    The criteria and methods used for the design of the control rods for the Experimental Gas-Cooled Reactor are described. The final mechanical design was derived from extensive thermal and mechanical calculations and actual experience obtained by fabrication of a prototype rod. The nuclear design of the rod was based on detailed calculations, the accuracy of which was checked by comparison with a measurement of rod worth made with the Physical Constants Test Reactor; By means of a meticulous application of basic principles the calculation agreed with the measurement within the experimental uncertainty. The most important nuclear aspect of the design is the large amount of epithermal absorption, which approximately doubles the worth over that of a purely thermal absorber. The rod is of an articulated type and consists of hot-pressed B4C-bushings clad in stainless-steel. The unique design of the load-supporting members allows operation at cladding temperatures up to 1600°F. Comparisons are made with control-rod designs for other gas-cooled reactors, and justifications for the choice of design features and material selection are discussed. The fabrication procedures and the final test programme for verification of the adequacy of the design are described. (author)

  18. 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

  19. Low-order dynamic modeling of the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    This report describes the development of a low-order, linear model of the Experimental Breeder Reactor II (EBR-II), including the primary system, intermediate heat exchanger, and steam generator subsystems. The linear model is developed to represent full-power steady state dynamics for low-level perturbations. Transient simulations are performed using model building and simulation capabilities of the computer software Matrixx. The inherently safe characteristics of the EBR-II are verified through the simulation studies. The results presented in this report also indicate an agreement between the linear model and the actual dynamics of the plant for several transients. Such models play a major role in the learning and in the improvement of nuclear reactor dynamics for control and signal validation studies. This research and development is sponsored by the Advanced Controls Program in the Instrumentation and Controls Division of the Oak Ridge National Laboratory. 17 refs., 67 figs., 15 tabs

  20. HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

    International Nuclear Information System (INIS)

    HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

  1. Experimental Breeder Reactor II (EBR-II), instrumentation for core surveillance

    International Nuclear Information System (INIS)

    The paper describes the Experimental Breeder Reactor-2 (EBR-2), thermal-hydraulic testing on the facility, and features of EBR-2 subassembly design. It is reported that during 25 years of EBR-2 operation, several of original, non-replaceable flow-sensors and thermocouples have failed in the primary system, and that this has led to the development of new sensors. The conclusion is made that from test series of measurements of temperature and flow in subassemblies, EBR-2 calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. 11 refs, 9 figs

  2. Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

  3. 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)

  4. Student Training Course Using the Experimental Fast Reactor JOYO and Related Facilities

    International Nuclear Information System (INIS)

    University level training courses have been initiated and implemented using the Experimental Fast Reactor Joyo and related facilities of the Japan Atomic Energy Agency (JAEA). These courses offer nuclear facility on-site education and experience in conjunction with a highly experienced engineering staff. University Nuclear Engineering Department faculty members have strongly supported and collaborated in the development of this program. The program covers reactor core physics analysis plus experiments using full-scope training simulator and performing neutron dosimetry, isotopic analysis of noble gases, chemical analysis of sodium, etc. This program is also anticipated to promote the human resource development in the younger generation for the nuclear industry, and to strengthen the relation between JAEA and University research programs. (author)

  5. Parametric and alternative studies for fusion experimental reactor (FER) (FY 1984)

    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. Starting from 1984 FER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. This report includes the following parametric and alternative studies for the FER reference design: 1) parametric studies concerning with core plasma magnets, and operation scenario and power supply, 2) tritium breeding blanket, 3) the study for the steady state operation FER, 4) OTHERS. (AUTHOR)

  6. 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)

  7. 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.

  8. 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

  9. Characterization of irradiation fields for fuel and material irradiation in the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    The Joyo MK-III core is a worldwide fast neutron irradiation field not only for FBR development but also for use in other fields such as light water reactor (LWR) and fusion reactor studies, and in the non-nuclear industry. The characterization of these neutron and gamma ray fields is most important to utilize for irradiation tests. This paper describes the details of distributions of neutron flux, reaction rate and gamma heating in the MK-III core. The calculation accuracy of the core management codes HESTIA, TORT and MCNP, was also evaluated by the measured data. The calculated results in neutron calculation agreed well with the measured one. The calculation method was validated and correction factors were identified. In case of gamma heating evaluation, the calculated result is underestimated with respect to the experimental value especially in the upper and lower SS reflector region. Further investigations in gamma heating evaluation are needed. (author)

  10. 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

  11. Radioactive waste treatment from experimental reactor PWR type developed to the Centro Experimental de Aramar -- Sao Paulo -- Brazil

    International Nuclear Information System (INIS)

    With the objective of getting experience on technology development to build PWR reactors in Brazil, a research center called Centro Experimental ARAMAR -- CEA was created. Taking into account all environmental factors, it had been necessary to develop a waste management program to meet this objective. This program, besides considering topics related to the safety and health of the public and worker, must consider the treatment of all radioactive waste (LLW and ILW) that will be generated, for minimizing the risks to the environment. COPESP has been developing compatible systems using technology through an extensive literature research about the state of the art in this field and practical experience from the Angra 1 Nuclear Power Plant. In this paper the authors will show some aspects related to the waste management program that have already been developed by COPESP, for instance, solids, liquids and gaseous radioactive waste management

  12. Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

    International Nuclear Information System (INIS)

    Full text: The work that has been conducted so far justifies a possibility of constructing a reactor with a non-traditional coolant to develop radically new reactors and their cycles with perfect architecture. A solid coolant, for example, the carbon-based one, allows to design the primary circuit of nuclear reactor without excess pressure. Such coolant withstands temperatures up to ∼4000 deg. K without a collapse. The analysis of theory and experiments produced requirements to be met by a solid coolant used in the primary circuit of nuclear reactor. One of the most important requirements is the arrangements for a continuous and homogeneous gravity flow of the coolant through all core sections taking into account the dust caused by wear and some amount of fractured particles. Therefore, the idea is that the mass of particles should resemble a liquid to a certain extend. The particles should be sphere like with average diameter from 0.5 to 2.0 mm and nonsphericity rate not more than 10%. 'Angle of repose' of particles to the horizon can be utilised as a validity criterion of particles which should not exceed 25 deg. The heat transfer coefficient should be increased up to the practical maximum value. In 1996 - 1997 the system of experimental facilities were built in the Scientific and Research Institute 'Luch' to prove the possibility to reliably cool a nuclear reactor with a flow of solid particles and to obtain a minimum set of data for the conceptual design of such reactor with solid coolant. The facility allows the research of the flow stability, heat mass transfer in the core, lifetime wearing of particles of the solid coolant. In 1994-1999 5 batches of particles of different size were fabricated in accordance to different technologies. Four batches were graphite-based and one was aluminium oxide-based (Al2O3). The purpose was to verify how the heat transfer coefficient was changing as the particle size varied. The average diameter of graphite particles was 0

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G.; Estrade, J. [French Atomic Energy Commission, Paris (France)

    2013-07-01

    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

  15. 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.

  16. Experimental modelling and numerical analysis of a molten salt fast reactor

    International Nuclear Information System (INIS)

    In this paper experimental and numerical investigation of the MSFR (Molten Salt Fast Reactor) concept will be presented. This homogeneous, single region liquid fuelled fast reactor concept uses fluoride-based molten salts with fissile uranium and thorium and other heavy nuclei content with the purpose of applying the thorium cycle and the burn-up of transuranic elements. Molten salt reactors with liquid fuel have a unique safety related property that needs clear understanding. In the core neutron flux and fission distribution is determined by the flow field through distribution and transport of fissile material and delayed neutron precursors. Since the MSFR concept has a single region homogeneous core without internal structures, it is a difficult task to ensure stable flow field, which is also strongly coupled to the volumetric heat generation. These considerations suggest that experimental and numerical modelling (including the option of coupled neutronics-thermal-hydraulics) would be needed to better understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built at BME NTI with the purpose of investigating the flow behavior inside the core region using particle image velocimetry. Not only the basic flow behavior inside the core region can be investigated but measurement data can also provide resource for the validation of computational fluid dynamics models, specific problems or phenomena (for example inlet geometry, optional internal structures, mixing) may be studied as well. Measurement results of steady state conditions will be presented with comparison of measurement data and results of numerical analyses. (author)

  17. The verifying test of refueling system of the China experimental fast reactor

    International Nuclear Information System (INIS)

    The article introduce the verifying test of refueling system of China Experimental Fast Reactor. The purposes of the test is to check the performance of the equipment of refueling system, and to verify the requirement for the SCADA (Supervisory Control and Data Acquisition) system, and to verify the refueling SCADA system. For these purposes the test platform and device were built. For the first time in China, the simulated automated refueling was realized on the platform. This test has established the base for the test of refueling system on CEFR. (authors)

  18. Renovation of new fuel transfer machine in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    In the higher performance plan (MK-III plan) of the experimental Fast Reactor JOYO, fuel handling system has been renovated to remote control system to reduce refueling time. As a part of this plan, new fuel transfer machine which is used to receive and transport new fuel, has been renovated completely to remote an automatic control system with no local operation and no local watching by Kawasaki Heavy Industries, Ltd. In this paper, the design and fabrication of this system are described. (author)

  19. Investigations into experimental steam generators for secondary water chemistry in pressurized water reactors

    International Nuclear Information System (INIS)

    Since 1981 investigations have been carried out on boiler models, which correspond in thermo-dynamic design to the boilers in nuclear power stations with Kraftwerk Union pressurized water reactors (PWR). The object of this experiment is to investigate in detail the effects of chemical operation in the secondary cycle of the PWR on the boilers and to test different conditioning procedures. Three test cycles have so far been undertaken over a three month period. The boilers have been operated using three different conditioning procedures, which were maintained for all cycles. This paper discusses the design of the experimental plant, its commissioning and the first three cycles. (orig.)

  20. A high-aspect-ratio design option for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Design features and performance estimates for HARD -- the high-aspect-ratio (A = 4) International Thermonuclear Experimental Reactor (ITER) design variant developed by the US ITER Team -- are presented. Key physics and engineering considerations associated with increased aspect ratio are described. The HARD design makes it possible for ITER to achieve both the ignition/extended-burn and steady-state/technology-testing performance goals set forth in ITER Terms of Reference. These capabilities are obtain in a device that is otherwise similar in concept, size and cost to the A = 2.8 ITER design defined by the Conceptual Design Activity

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

    International Nuclear Information System (INIS)

    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 deg. C. The rate of the conversion from UN to U3O8 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

  2. 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.

  3. Design study of the FER (fusion experimental reactor) superconducting magnet system

    International Nuclear Information System (INIS)

    The preliminary design study of the superconducting magnet system for the Fusion Experimental Reactor (FER) [1] is in progress at the Japan Atomic Energy Research Institute (JAERI). This paper describes the technical specifications and the design concepts of the 12-T toroidal field (TF) coil system, the 12-T central solenoid (CS) coil system, and the 7-T equilibrium field (EF) coil system, all of which satisfy the technical requirements of the superconducting magnet system of the FER. 8 refs., 6 figs., 4 tabs

  4. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report

    International Nuclear Information System (INIS)

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D2O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period

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

    International Nuclear Information System (INIS)

    Highlights: ► Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. ► Vibroacoustic investigations. ► Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. ► 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 °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 electromagnetic bearings system operability.

  6. 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

  7. CARINA. A program for experimental investigation of the irradiation behaviour of German reactor pressure vessel materials

    International Nuclear Information System (INIS)

    The proof of a sufficient safety margin against brittle fracture of the reactor pressure vessel (RPV) is an important part of the operational safety of nuclear power plants. The RPV safety assessment procedure applicable in Germany is described in KTA 3201.2 of the Nuclear Safety Standard Commission (KTA). This deterministic assessment concept is based on the comparison of load curves with the material resistance curve in terms of fracture toughness. The fracture toughness curve can be determined either indirectly according to the RTNDT concept based on Charpy tests or directly according to the more appropriate RTT0 approach based on Master Curve analysis of fracture toughness tests, respectively. In the recently completed research project CARINA the data base for pre-irradiated original RPV steels of German PWR construction lines was extended by comprehensive fracture toughness testing. The data obtained up to neutron fluences of 7.67 x 1019 n/cm2 (E > 1 MeV) are analysed and discussed particularly in terms of Master Curve applications. The experimental results show that optimized RPV manufacturing specifications and reactor designs are advantageous for a long-term plant operation in comparison to less optimized materials with lower toughness and to reactor designs with substantial higher neutron irradiation. With the obtained data, experiences and insights an essential contribution was also made to the integration of the Master Curve concept in German safety standards.

  8. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    International Nuclear Information System (INIS)

    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 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. 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

  9. Experimental irradiations of research reactor fuels and some post-irradiation subjects

    International Nuclear Information System (INIS)

    Full text: This work was performed by request of the Fuel Elements Department, in relation to techniques usually used in post-irradiation examinations and to the actual availability of equipment and specialists for the implementation of those techniques. Since the specialists are working in different areas of National Atomic Energy Commission (CNEA), a data base is formulated to include this knowledge, which will also help to take decisions. Besides, in relation to chemical analysis, a tree-shaped decision procedure is described, which enables, given an irradiated material sample, an orientation of the following stages, in order to obtain the results required according to the formulated irradiation experiment. Furthermore, the design of an equipment for fission gas collection is shown. Regarding the possible irradiation experiences to be performed in the experimental reactor, an experience was formulated to determine densification, an important parameter to measure, in order to obtain relevant information on fuel engineering, since it directly influences on the powder and pellet fabrication methodology. As a higher flux reactor is not available, the way to irradiate capsules in the RA-3 reactor is suggested in order to obtain irradiated fuel to be analyzed in hot cells. The convenience to buy external irradiation information for code syntonization and qualification, as well as for training is discussed, a local infrastructure might be built to obtain basic and engineering information for the development of new fuels

  10. CARINA. A program for experimental investigation of the irradiation behaviour of German reactor pressure vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Hein, Hieronymus; Keim, Elisabeth [AREVA GmbH, Erlangen (Germany); Bechler, Eduard [E.ON Kernkraftwerk GmbH, Hannover (Germany); Efsing, Paal [Ringhals AB, Vaeroebacka (Sweden); Ganswind, Jens [VGB PowerTech e.V., Essen (Germany); Knobel, Rene [Kernkraftwerk Goesgen-Daeniken AG, Daeniken (Switzerland); Koenig, Guenter [EnBW Kernkraft GmbH, Kernkraftwerk Neckarwestheim, Neckarwestheim (Germany); Barreiro, Pablo [EnBW Kernkraft GmbH, Kernkraftwerk Philippsburg, Philippsburg (Germany); Widera, Martin [RWE Power AG, Essen (Germany); Jong, Andre de [N.V. EPZ Kerncentrale Borssele, Borssele (Netherlands)

    2013-05-15

    The proof of a sufficient safety margin against brittle fracture of the reactor pressure vessel (RPV) is an important part of the operational safety of nuclear power plants. The RPV safety assessment procedure applicable in Germany is described in KTA 3201.2 of the Nuclear Safety Standard Commission (KTA). This deterministic assessment concept is based on the comparison of load curves with the material resistance curve in terms of fracture toughness. The fracture toughness curve can be determined either indirectly according to the RTNDT concept based on Charpy tests or directly according to the more appropriate RTT0 approach based on Master Curve analysis of fracture toughness tests, respectively. In the recently completed research project CARINA the data base for pre-irradiated original RPV steels of German PWR construction lines was extended by comprehensive fracture toughness testing. The data obtained up to neutron fluences of 7.67 x 10{sup 19} n/cm{sup 2} (E > 1 MeV) are analysed and discussed particularly in terms of Master Curve applications. The experimental results show that optimized RPV manufacturing specifications and reactor designs are advantageous for a long-term plant operation in comparison to less optimized materials with lower toughness and to reactor designs with substantial higher neutron irradiation. With the obtained data, experiences and insights an essential contribution was also made to the integration of the Master Curve concept in German safety standards.

  11. 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)

  12. 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.

  13. 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

  14. Major accident analyses for experimental zero-power fast reactor assemblies

    International Nuclear Information System (INIS)

    A study has been made of the possibility, mechanism, and consequence of melt-down and other major nuclear accidents for a ZPR-III type experimental zero-power fast reactor of the two-half type. This study has been supplemented by an evaluation of the importance of the Doppler effect for a wide range of nuclear reactor assemblies for such a reactor. A melt-down event is highly improbable because of the restricted sequence of events which must be postulated. A discussion of the mechanism of the collapse is followed by the results of coupled neutronics-hydrodynamic s calculations for two zero-power assemblies. A 1200-l core has been examined because it represents a relatively large reactor of common core composition. A smaller core with a high-void fraction has been examined as a potentially more dangerous system. Very different time-wise behaviour has been found for the two systems. For sharp accidents in zero-power assemblies, the U235-atoms, separated as plates of enriched uranium, will heat very rapidly while the remainder of the core remains essentially cold, so that a gas of U235-vapour will provide the disassembly pressure. The adaption of the neutronics-hydrodynamic s code AX-I to the use of a Van der Waals gas is described. Another important change in the equation of state used in the code is to employ a Mie-Griineisen type equation derivable from solid state theory. This change provides a more satisfactory way to evaluate the pressure term for cores of variable composition. Because the highly enriched U235 plates of a zero-power assembly will heat much more rapidly than the depleted uranium plates, the possibility of a net positive Doppler effect is much larger for an experimental assembly than for the equivalent power breeder reactor. This hazard has been examined for a range of possible assemblies. These calculations indicate that the Doppler coefficient for a zero-power assembly does not become important as a hazard until one approaches systems with the

  15. Upgrade of Cooling System Heat Removal Capacity of the Experimental Fast Reactor JOYO

    International Nuclear Information System (INIS)

    The purpose of the MK-III program is to upgrade the irradiation capability of the liquid sodium-cooled experimental fast reactor JOYO. As a result, the neutron flux density of the core was increased, and the reactor thermal power was increased to 140 MW(thermal) from the originally designed 100 MW(thermal). To accommodate the increased thermal power, the flow rates of sodium coolant in the primary and secondary systems were increased by 20 and 10%, respectively. Also, all intermediate heat exchangers and dump heat exchangers were replaced with new ones. The replacement of these large sodium components was carried out over an [approximately]1-yr period with both fuel and molten sodium still in the reactor vessel (RV).Major challenges in the replacement were the control of impurity ingress to existing systems and protection from radiation exposure in the high-dose-rate regions inside the containment vessel. During the replacement, the seal bag method, impurity concentration monitoring of cover gas, and low-pressure control of cover gas were applied to prevent damage to existing components and systems, such as the RV, fuel subassemblies, sodium piping, and tanks. The measures taken to reduce the radiation exposure were a lowering of the surrounding dose rate through the use of temporary shielding, shortening of the operation time near the high-dose-rate area by first doing thorough training, and the employment of protection equipment to avoid contamination. The replacement of components was completed without major trouble, and methods applied for the replacement proved to be effective in the operation and maintenance of sodium-cooled reactors

  16. Decontamination and decommissioning of the Experimental Boiling Water Reactor at Argonne National Laboratory

    International Nuclear Information System (INIS)

    The Experimental Boiling Water Reactor (EBWR), located on the Argonne National Laboratory-East (ANL-E) site, started operations in 1957. The initial rating was 20 MW(t). The rating was eventually increased to 70 MW(t) in 1959 and 100 MW(t) in 1962. The reactor was shut down in 1967 and all of the fuel was removed from the facility. The facility was placed in dry lay-up until 1986. ANL-E personnel started the decontamination and decommissioning (D ampersand D) effort in 1986. Supporting equipment such as the external steam system and some of the upper reactor components, the core riser and the top fuel shroud, were removed at that time. Characterization of the facility was also undertaken. The contract to complete the EBWR D ampersand D Project was issued in December 1993. The initial schedule called for the final effort to be divided into five phases that were to be completed over a four year period. However, this schedule was subsequently consolidated, at the request of ANL-E, to a thirteen month period, with the on-site work to be completed by the end of 1994. The EBWR D ampersand D Project is approximately 88% complete. A small quantity of reactor internals remains to be volume reduced along with the removal of the SFSP water treatment system. Upon completion of this work the facility will be decontaminated and a final survey completed. The planned completion of on-site work is scheduled for July 1995

  17. 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.)

  18. 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.

  19. Methods and experimental coefficients used in the computation of reactor shielding

    International Nuclear Information System (INIS)

    1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)

  20. 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

  1. FLICA III. A digital computer program for thermal-hydraulic analysis of reactors and experimental loops

    International Nuclear Information System (INIS)

    This computer program describes the flow and heat transfer in steady and transient state in two-phase flows. It is the present stage of the evolution about FLICA, FLICA II and FLICA II B codes which have been used and developed at CEA for the thermal-hydraulic analysis of reactors and experimental loops with heating rod bundles. In the mathematical model all the significant terms of the fundamental hydrodynamic equations are taken into account with the approximations of turbulent viscosity and conductivity. The two-phase flow is calculated by the homogeneous model with slip. In the flow direction an implicit resolution scheme is available, which make possible to study partial or total flow blockage, with upstream and downstream effects. A special model represents the helical wire effects in out-of pile experimental rod bundles

  2. Experimental tests and calculation methods for missile crashing effects on a reactor containment

    International Nuclear Information System (INIS)

    In the analysis of missile crashing on a reactor containment there are two main effects to be taken into account: the overall behaviour of the building; the local perforation. The overall behaviour of the building is easily calculated when the applied force as a function of time is known. Two calculation examples are presented. The local perforation is a much more difficult problem and experimental work is necessary. The report presents a series of perforation tests of concrete plates by cylindrical missiles with a flat nose. The aim of these tests is to extrapolate for the lower speeds the existing experimental correlations and to check the calculation methods. The calculations are made with the PASTEL code (Finite elements, implicit integration), with elastoplasticity of the reinforcing steel bars and the concrete. Various plastification and fracturation laws are tested. (Auth.)

  3. 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

  4. 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)

  5. Student internship program using the experimental fast reactor Joyo and related facilities

    International Nuclear Information System (INIS)

    The student training courses using the experimental fast reactor Joyo of the Japan Atomic Energy Agency (JAEA) and related facilities have been initiated based on the JAEA's mission to contribute to the human resources development program of the Japanese Ministry of Education, Culture, Sports, Science and Technology (MEXT) and Ministry of Economy, Trade and Industry (METI). The development of the student training courses was also strongly supported by the faculty of nuclear engineering of domestic universities in two reasons: one is that the nuclear related curriculum has recently been reduced due to the trend of decreasing interest by the younger generation in nuclear research and industry, and the other reason is that the aging research reactors and nuclear facilities owned by the universities are very difficult to keep operating. Considering this situation, JAEA decided to cooperate with the universities in developing the student training course. The experimental fast reactor Joyo of JAEA is a sodium cooled fast reactor with plutonium- uranium mixed oxide (MOX) fuel, which has two primary sodium loops, two secondary loops, and an auxiliary system. An intermediate heat exchanger (IHX) separates radioactive sodium in the primary system from non-radioactive sodium in the secondary system. The secondary sodium loop transports the reactor heat from the IHX to the air-cooled dump heat exchanger (DHX). Joyo has a full-scope type core and plant simulator, which duplicated all the main control panels located in the Joyo central control room. The simulator enables to offer a real time simulation of the plant behaviors under normal and abnormal conditions by applying the plant dynamic analysis code Minir-N2 and the same interlock system as the Joyo reactor system. The sodium analysis facility is located apart from Joyo complex to primarily conduct impurity measurement of Joyo cooling system. These data were measured by chemical analysis, gas chromatography, beta

  6. The AKR training reactor of the University of Technology Dresden and its experimental programme for education

    International Nuclear Information System (INIS)

    The training and research reactor AKR (from the German Ausbildungskernreaktor) of the University of Technology Dresden was put into operation in the late 1970s, i.e. at a time when experts in Western and Eastern European countries expected an extensive development of nuclear energy generation. In the eastern part of Germany, i.e. the former German Democratic Republic, a 70 MW(el) WWER nuclear power plant (NPP) was in operation at Rheinsberg since 1966. A further five NPPs of the Russian WWER type (440 MW(e)) were put into operation at Greifswald on the Baltic Sea coast between 1973 and 1989. Three more plants were under construction at that site. At the end of the 1980s, a new NPP was being planned at Stendal, a 1000 MW(e) WWER plant. For construction, licensing, operation and maintenance of these NPPs, a well educated and trained staff was required. Authorities and technical surveillance organizations should have qualified experts at their disposal, too. Consequently, in 1968 the University of Technology Dresden introduced appropriate courses for the education of nuclear engineers in order to contribute to fulfilling the demands of industry, science and administration. Students were taught in lectures, but theoretical knowledge had to be combined with practical experience based on an extensive programme of fundamental experiments in the fields of reactor physics, neutron physics, nuclear technology, radiation measurement techniques, radiation protection, radiation dosimetry and others. The full scale of this experimental programme can preferably be made available by small training reactors which can be operated with great diversity in terms of experimental intentions and without commercial restrictions

  7. 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

  8. Utilization of the experimental reactor Osiris for the study and the development of fuels of the fast neutron reactor type

    International Nuclear Information System (INIS)

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation

  9. 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)

  10. Radioprotection problems resulting from the presence of experimental devices around an atomic reactor

    International Nuclear Information System (INIS)

    The setting up of experimental devices around a reactor produces dangers of irradiation and radioactive contamination which can become very great in the case of an accident, especially if the in-pile portion contains fissile matter. This may result in irradiation of personnel, prohibition of access to the experimental zones until the sources of irradiation and contamination have been eliminated, and a prolonged stoppage of the reactor. The plans for an in-pile experiment should take into account radioprotection factors; the aim of these is to reduce to a minimum the radioactive risks normally encountered during the experiment and to eliminate any risks of bad accidents and their consequences. In this report are classified the various types of experiments requiring installations outside the pile itself; for each of these experiments the particular radioprotection factors are given. In order to make possible a study of the radioactive dangers likely to arise during a projected experiment, the authors summarize the physical and technical data required by radioprotection specialists and give the rules and general advice concerning radioprotection which should be useful during the planning of an in-pile experiment and the setting-up of the equipment. (authors)

  11. A new concept of laser fusion experimental reactor with fast ignition target

    International Nuclear Information System (INIS)

    Full text: We have analyzed the design windows of laser fusion power plants based on fast ignition targets, and examined feasibility of a small-sized laser fusion experimental reactor suitable for developing their power plants. Target gain curves are evaluated for power plants, which have 100∼200MJ fusion yields with 600kJ∼1MJ lasers, and for an experimental reactor (LFER), which has a 10MJ fusion yield with a 200kJ laser, 100kJ for implosion and 100kJ for heating. The pulse heat loads on the chamber wall of LFER are estimated at 2.5J/cm2 for a 2.5-m-radius solid wall chamber, and 16J/cm2 for a 1-m-radius liquid wall chamber. The fast ignition LFER can make its fusion output one order smaller than that of the central ignition, thus we can use a rather small solid wall chamber for the first stage of the LFER. We can also expect to decrease laser cost drastically, although for a heating laser we must develop the long life final optics. Through a fast ignition LFER, we showed a possibility to demonstrate net electric generation in a reasonably short time. (author)

  12. Experimental study on circulation characteristics of secondary passive heat removal system for Chinese pressurized water reactor

    International Nuclear Information System (INIS)

    Nuclear safety has attracted increasingly global attention. The secondary side natural circulation heat removal system is one of several passive safety systems designed to ensure the safety of the reactor core. The passive heat removal systems are also adopted in the safety system design of Chinese third-generation pressurized water reactor. To complete its design and verify its feasibility, an experimental facility, which is named SPHRS, was built to investigate the heat removal capacity and the stability characteristics of the steam generator secondary side passive heat removal system. The experimental facility consists of a steam generator simulator, an instrumented condenser test section with a secondary pool boiling section, a pump, a water storage tank, and associated piping and instrumentation. The heat removal capacity under varying pressures and characteristics of natural circulation at steady-state as well as system power input drop condition were obtained. The results show that the SPHRS has sufficient capacity to remove the decay heat. It is found that the natural circulation can be maintained under different pressures, the heat load sudden change condition as well, which indicates that the SPHRS natural circulation has good stability. - Highlights: • The stability of secondary side passive heat removal system of a PWR was studied. • The heat removal capacity of the condenser tube was investigated. • The natural circulation characteristics of SPHRS were investigated

  13. 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.

  14. A new concept of Laser Fusion Experimental Reactor with fast ignition target

    International Nuclear Information System (INIS)

    We have analyzed the design windows for laser fusion power plants based on fast ignition concepts, and examined the feasibility of a small-sized laser fusion experimental reactor suitable for developing their power plants. Target gain curves are assessed for power plants, having 90∼200 MJ fusion yields with 600 kJ∼1MJ lasers, and for an experimental reactor (LFER), having a 10 MJ fusion yield with a 200 kJ laser, i.e., 100 kJ for implosion and 100 kJ for heating. The pulse heat loads on the chamber wall of LFER are estimated as 2.5 J/cm2 for a 2.5-m-radius solid wall chamber, and 16 J/cm2 for a 1-m-radius liquid wall chamber. The fast ignition LFER can produce its fusion output approximately one order of magnitude smaller than that of the central ignition, so that we can use a rather small solid wall chamber for the first stage of the LFER operation. We can also expect to decrease laser cost drastically, although for the heating laser we must develop a long life final optics system. With the fast ignition LFER, we showed a possibility to demonstrate net electric generation in a reasonably short time. (author)

  15. Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

    International Nuclear Information System (INIS)

    A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

  16. 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)

  17. 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

  18. Experimental studies of some of the physical features of beryllium-moderated intermediate reactors

    International Nuclear Information System (INIS)

    This paper is devoted to a review of the results obtained from a number of experiments carried out on the PF-4 critical assembly (intermediate-physical assembly), which is designed for detailed studies of the physical characteristics of intermediate reactors. The cores and reflectors of the various critical assemblies were comprised of compact units of steel or aluminium tubes, in which discs of various materials were placed. By combining 90%-enriched uranium discs with moderating materials in various proportions, and also by introducing moderating layers of different thicknesses into the reflector, it was possible to alter the spectrum of the fission-inducing neutrons within a very broad range. This paper describes the PF-4 critical assembly and its various subassemblies. The relative efficiency of internal and external moderation is analysed for reactors with a very low ratio of moderator nuclei to uranium nuclei in the core. The experiments show that even when thick moderating reflectors are used, this low ratio (the ratio of beryllium nuclei to uranium-235 nuclei being ∂Be/∂U235≅1) leads to an increase of the critical mass. A considerable part of the paper is devoted to an analysis of heterogeneous effects in intermediate reactors. It is shown that for reactors with ∂Be/∂U235=30-40 various thicknesses of highly enriched uranium, ranging from 0.023 g/cm2 to 32 g/cm2, have an identical effect on the reactivity of the system. The causes underlying compensation of the neutron-flux screening effect by thick layers of uranium are analysed. The interesting fact that the efficiency of uranium increases in the neighbourhood of the absorbing rods, which was experimentally revealed in an assembly with ∂Be/∂U235≅200, is discussed in the paper. This fact is explained by the sharp decline in the importance of neutrons absorbed by the uranium. The paper provides data, derived from the same assembly, on the efficiency of rods made of various absorbing materials

  19. Experimental and Modeling Studies of the Methane Steam Reforming Reaction at High Pressure in a Ceramic Membrane Reactor

    OpenAIRE

    Hacarlioglu, Pelin

    2007-01-01

    This dissertation describes the preparation of a novel inorganic membrane for hydrogen permeation and its application in a membrane reactor for the study of the methane steam reforming reaction. The investigations include both experimental studies of the membrane permeation mechanism and theoretical modeling of mass transfer through the membrane and simulation of the membrane reactor with 1-D and 2-D models. A hydrothermally stable and hydrogen selective membrane composed of silica and a...

  20. 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

  1. WWER-440 power peaking experiment with/without Hf inserts in the LR-0 reactor. MCNP calculations vs. experimental results

    International Nuclear Information System (INIS)

    In this presentation authors deal with power peaking experiment with/without Hf inserts in the LR-0 reactor. Obtained experimental results are compared with MCNP calculations. The MCNP model gives accurate fission density predictions (not XS data sensitive). The criticality calculation depends significantly on the XS library. The experiments in the LR-0 reactor provide reproducible high quality data. This data will be used to define a set of benchmarks for industrial and licensing needs.

  2. Theoretical and experimental modeling of the multiple pressure tube rupture for RBMK reactor. Pt. 1

    International Nuclear Information System (INIS)

    Rupture of a single fuel channel (pressure tube) or several fuel channels of the RBMK may occur in service conditions on NPPs with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighboring channels to break. This assumption has not been justified. Hence, an analysis of the multiple pressure tube rupture (MPTR) possibility is needed. The analysis of the MPTR problem requires performing a series of theoretical and experimental studies of separate physical processes running in the RBMK reactor, as well as development of mathematical models and their physical equivalents. The experimental rigs concerned the MPTR problem have been designed and constructed at Electrogorsk Research and Engineering Center, Russia. Investigation of the circumstances and mechanisms of a single channel rupture at the various conditions and scenarios is one of the main stages of the MPTR problem analysis. Theoretical models of the single channel rupture under thermal and mechanical loading have been developed including a channel constrained by the graphite block. Deformation of the channel under internal pressure and localized thermal action is modeled within the framework of the nonlinear shells theory taking into account physically nonlinear material behavior. Computer program based on these models enables to describe the thermomechanical deformation of a single channel and to predict rupture moment. Theoretical studies were accompanied by experimental modeling single channel rupture by means of series experimental examinations at TKR-F test rig (Model of an Accidental Channel). This test rig represents a model of the single disrupted fuel channel in a surrounding graphite column. Experimental examinations make possible the development and verification of theoretical models and make more exact the conditions and mechanism of a single channel rupture. Theoretical and experimental modeling consolidation sets out technique

  3. 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

  4. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

    International Nuclear Information System (INIS)

    An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

  5. 04 - Sodium cooled fast breeder fourth-generation reactors - The experimental reactor ALLEGRO, the other ways for fast breeder fourth-generation reactors

    International Nuclear Information System (INIS)

    The authors first present the technology of gas-cooled fast breeder reactors (basic principles, specific innovations, feasibility studies, fuel element, safety) and notably the ALLEGRO project (design options and expected performances, preliminary safety demonstration). Then, they present the lead-cooled fast-breeder reactor technology: interests and obstacles, return on experience, the issue of lead density, neutron assessment, transmutation potential, dosimetry, safety chemical properties and compatibility with the fuel, water, air and steels. The next part addresses the technology of molten-salt fast-breeder reactors: choice of the liquid fuel and geometry, reactor concept (difficulties, lack of past R and D), demonstration and demonstrators, international context

  6. Experimental study on the disinfection efficiencies of a continuous-flow ultrasound/ultraviolet baffled reactor.

    Science.gov (United States)

    Zhou, Xiaoqin; Guo, Hao; Li, Zifu; Zhao, Junyuan; Yun, Yupan

    2015-11-01

    A self-designed continuous-flow ultrasound/ultraviolet (US/UV) baffled reactor was tested in this work, and the disinfection efficiency of secondary effluent from a wastewater treatment plant (WWTP) was investigated in terms of the different locations of ultrasonic transducers inside the reactor under similar input power densities and specific energy consumptions. Results demonstrated that the two-stage simultaneous US/UV irradiation in both chambers 2 and 3 at a flow rate of 1200 L/h performed excellent disinfection efficiency. It achieved an average feacal coliforms concentration of 201±78 colony forming unit (CFU)/L in the effluent and an average of (4.24±0.26) log10 reduction. Thereafter, 8 days of continuous operation was performed under such a condition. A total of 31 samples were taken, and all the samples were analyzed in triplicate for feacal coliforms analysis. Experimental results showed that feacal coliforms concentrations remained at about 347±174 CFU/L under the selected optimum disinfection condition, even if the influent concentrations fluctuated from 3.97×10(5) to 3.57×10(6) CFU/L. This finding implied that all effluents of continuous-flow-baffled-reactor with simultaneous US/UV disinfection could meet the requirements of the discharge standard of pollutants for municipal WWTP (GB 18918-2002) Class 1-A (1000 CFU/L) with a specific energy consumption of 0.219 kWh/m(3). Therefore, the US/UV disinfection process has great potential for practical applications. PMID:26186823

  7. 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

  8. 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.)

  9. Development of Experimental System for Material Compatibility Test for Ultra-long Cycle Fast Reactor (UCFR)

    International Nuclear Information System (INIS)

    Sodium is a candidate for fast reactor coolants that has been believed to have favorable compatibility with structural materials. However, recent studies showed results which need for a more careful attention at this previous belief. For prolonging the service life time of cladding and structural materials in contact with liquid sodium, more detail analysis methods are needed to examine this material compatibility issue with sodium. As a candidate of liquid metals coolants of Ultra-long Cycle Fast Reactor (UCFR), the compatibility of sodium with cladding materials has to be investigated in detail with long term exposure time. It is known that sodium promotes corrosion in two ways. One is corrosion produced by dissolution of alloy elements into sodium and the other is corrosion produced through a chemical reaction with impurities in sodium (especially, dissolved oxygen). The use of the technique of impedance spectroscopy to measure the electrical impedance response of any oxide layers may be a good experimental tool to this monitoring system. The motivation of current study is to investigate the relationship between the electrochemical behaviors of oxide scales on martensitic and austenitic steels and their corrosion rates in liquid sodium

  10. Experimental study of flow inversion in MTR upward flow research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Hadi, Ead A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; Khedr, Ahmed; Talha, Kamal Eldin Aly; Abdel-Latif, Salwa Helmy

    2014-06-15

    The core cooling of upward flow MTR pool type Research Reactor (RR) at the later stage of pump coast down is experimentally handled to clarify the effect of some operating parameters on RR core cooling. Therefore, a test rig is designed and built to simulate the core cooling loop at this stage. The core is simulated as two vertical channels, electrically heated, and extended between upper and lower plenums. Two elevated tanks filled with water are connected to the two plenums. The first one constitutes a left branch, connected to the lower plenum, and is electrically heated to simulate the core return pipe. The second one constitutes the right branch, connected to the upper plenum, and is cooled by refrigerant circuit to simulate the reactor pool. Channel coolant and wall temperatures at different power and branch temperatures are measured, registered and analyzed. The results show that at this stage of core cooling two cooling loops are established; an internal circulation loop between the channels dominated by the difference in channel's power and an external circulation loop between the branches dominated by the temperature difference between branches. Also, there is a double inversion in core flow, upward-downward-upward flow. This double inversion increases largely the channel's wall temperature. Complementary safety analysis to evaluate this phenomenon must be performed. (orig.)

  11. Experimental study of flow inversion in MTR upward flow research reactors

    International Nuclear Information System (INIS)

    The core cooling of upward flow MTR pool type Research Reactor (RR) at the later stage of pump coast down is experimentally handled to clarify the effect of some operating parameters on RR core cooling. Therefore, a test rig is designed and built to simulate the core cooling loop at this stage. The core is simulated as two vertical channels, electrically heated, and extended between upper and lower plenums. Two elevated tanks filled with water are connected to the two plenums. The first one constitutes a left branch, connected to the lower plenum, and is electrically heated to simulate the core return pipe. The second one constitutes the right branch, connected to the upper plenum, and is cooled by refrigerant circuit to simulate the reactor pool. Channel coolant and wall temperatures at different power and branch temperatures are measured, registered and analyzed. The results show that at this stage of core cooling two cooling loops are established; an internal circulation loop between the channels dominated by the difference in channel's power and an external circulation loop between the branches dominated by the temperature difference between branches. Also, there is a double inversion in core flow, upward-downward-upward flow. This double inversion increases largely the channel's wall temperature. Complementary safety analysis to evaluate this phenomenon must be performed. (orig.)

  12. 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

  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. 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

  15. Methodology for the ITER [International Thermonuclear Experimental Reactor] technology phase operational scenario analysis

    International Nuclear Information System (INIS)

    The operational space for the technology phase of the International Thermonuclear Experimental Reactor (ITER) has been examined with the ITER systems code, TETRA (September 1989 version). The goal of the technology phase is to provide neutron wall loads sufficient to test the technologies required for fusion reactors, ideally with steady-state operation. Heat loads on the divertor, however, place severe restrictions on the operational space. Steady-state operation is found to be limited to neutron wall loads of about 0.75 MW/m2 (Q near 6, and injection power of 150 MW), even when controlled impurity seeding is used to mitigate the divertor heat loads via line radiation. Wall loads approaching the required levels for technology testing (∼1 MW/m2) are possible with steady-state operation only if the divertor limits are relaxed. Otherwise, a hybrid form of operation, in which both inductive and noninductive current drive are used, must be employed for technology testing. With this scheme, plasma operational space is found with long pulses (>1000 s), high wall loads (1.2 MW/m2), energy multiplication factor Q > 10, injection power near 100 MW, and acceptable divertor conditions. We note that the results for ITER operation have changed slightly since this work was carried out. Nonetheless the basic methodology and the trends identified here remain the same. 16 refs

  16. Experimental validation of photon-heating calculation for the Jules Horowitz Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lemaire, M., E-mail: matthieu.lemaire@cea.fr [CEA, DEN, DER, SPRC, Cadarache, F-13108 Saint Paul lez Durance (France); Vaglio-Gaudard, C. [CEA, DEN, DER, SPRC, Cadarache, F-13108 Saint Paul lez Durance (France); Lyoussi, A. [CEA, DEN, DER, SPEx, Cadarache, F-13108 Saint Paul lez Durance (France); Reynard-Carette, C. [Aix Marseille Université, CNRS, Université de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Di Salvo, J.; Gruel, A. [CEA, DEN, DER, SPEx, Cadarache, F-13108 Saint Paul lez Durance (France)

    2015-04-21

    The Jules Horowitz Reactor (JHR) is the next Material-Testing Reactor (MTR) under construction at CEA Cadarache. High values of photon heating (up to 20 W/g) are expected in this MTR. As temperature is a key parameter for material behavior, the accuracy of photon-heating calculation in the different JHR structures is an important stake with regard to JHR safety and performances. In order to experimentally validate the calculation of photon heating in the JHR, an integral experiment called AMMON was carried out in the critical mock-up EOLE at CEA Cadarache to help ascertain the calculation bias and its associated uncertainty. Nuclear heating was measured in different JHR-representative AMMON core configurations using ThermoLuminescent Detectors (TLDs) and Optically Stimulated Luminescent Detectors (OSLDs). This article presents the interpretation methodology and the calculation/experiment (C/E) ratio for all the TLD and OSLD measurements conducted in AMMON. It then deals with representativeness elements of the AMMON experiment regarding the JHR and establishes the calculation biases (and its associated uncertainty) applicable to photon-heating calculation for the JHR.

  17. Experimental validation of photon-heating calculation for the Jules Horowitz Reactor

    Science.gov (United States)

    Lemaire, M.; Vaglio-Gaudard, C.; Lyoussi, A.; Reynard-Carette, C.; Di Salvo, J.; Gruel, A.

    2015-04-01

    The Jules Horowitz Reactor (JHR) is the next Material-Testing Reactor (MTR) under construction at CEA Cadarache. High values of photon heating (up to 20 W/g) are expected in this MTR. As temperature is a key parameter for material behavior, the accuracy of photon-heating calculation in the different JHR structures is an important stake with regard to JHR safety and performances. In order to experimentally validate the calculation of photon heating in the JHR, an integral experiment called AMMON was carried out in the critical mock-up EOLE at CEA Cadarache to help ascertain the calculation bias and its associated uncertainty. Nuclear heating was measured in different JHR-representative AMMON core configurations using ThermoLuminescent Detectors (TLDs) and Optically Stimulated Luminescent Detectors (OSLDs). This article presents the interpretation methodology and the calculation/experiment (C/E) ratio for all the TLD and OSLD measurements conducted in AMMON. It then deals with representativeness elements of the AMMON experiment regarding the JHR and establishes the calculation biases (and its associated uncertainty) applicable to photon-heating calculation for the JHR.

  18. Experimental validation of photon-heating calculation for the Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is the next Material-Testing Reactor (MTR) under construction at CEA Cadarache. High values of photon heating (up to 20 W/g) are expected in this MTR. As temperature is a key parameter for material behavior, the accuracy of photon-heating calculation in the different JHR structures is an important stake with regard to JHR safety and performances. In order to experimentally validate the calculation of photon heating in the JHR, an integral experiment called AMMON was carried out in the critical mock-up EOLE at CEA Cadarache to help ascertain the calculation bias and its associated uncertainty. Nuclear heating was measured in different JHR-representative AMMON core configurations using ThermoLuminescent Detectors (TLDs) and Optically Stimulated Luminescent Detectors (OSLDs). This article presents the interpretation methodology and the calculation/experiment (C/E) ratio for all the TLD and OSLD measurements conducted in AMMON. It then deals with representativeness elements of the AMMON experiment regarding the JHR and establishes the calculation biases (and its associated uncertainty) applicable to photon-heating calculation for the JHR

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

    International Nuclear Information System (INIS)

    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

  20. 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

  1. 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.

  2. 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.

  3. Experimental demonstration of proportional-integral-derivative feedback in the closed-loop digital control of reactor neutronic power

    International Nuclear Information System (INIS)

    This paper reports the design, implementation, and experimental evaluation of a proportional-integral-derivative (P-I-D) feedback methodology for the closed-loop, digital control of neutronic power in a nuclear reactor. The approach used is innovative in that each of the control actions that comprise the error signal is chosen so as to have the mathematical form of a reactor period. The resulting error signal is then implemented using the MIT-SNL Period-Generated Minimum Time Control Laws which incorporate a non-linear model of the reactor's dynamics. This feedback methodology was assessed and shown to be effective during the course of fifty-six experimental runs, all performed under conditions of closed-loop, digital control. In addition to describing this P-I-D methodology and the experimental results, a brief review is given of both the MIT-SNL laws and the theory of P-I-D control

  4. 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.

  5. Replacement of secondary heat transport system components in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    A recently completed major upgrade of the JOYO experimental sodium-cooled fast reactor, to the MK-III design, increased its irradiation capability. One major change was a 40% increase in thermal power to 140 MWt, which necessitated the replacement of the cooling system. Major challenges in the replacement of secondary components were control of impurity ingress and assurance of welding integrity. Damage to existing systems was avoided during replacement operations by taking measures to prevent ingress of air into the sodium systems. The long exposure of the used pipes made of ferritic low-alloy steel to hot sodium was a concern because previous research showed that this material changes its mechanical property in sodium hotter than 673 K. Used pipe, heat transfer tubes and welds were subjected to material tests. These tests did not show notable material problems. The replacement of components was completed without major troubles, demonstrating the effectiveness of the methods used. (authors)

  6. 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

  7. 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)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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

  13. Modular gas reactor lift-off source term: data needs and experimental plans

    International Nuclear Information System (INIS)

    The Modular Gas-cooled Reactor incorporates many features which could make it attractive to US utilities. Foremost among these is enhanced safety, achieved through simplified design and reliance on passive safety features. For the MGR to meet its goal of not requiring public evacuation even during the design basis depressurization accident may require either a reduction in the uncertainty in source term calculations or extremely high fuel quality. A major source of this uncertainty is in the prediction of deposition and lift-off of radioactive contaminants on primary circuit walls. Previous experiments have not produced data which is readily usable for MGR design studies. MIT's DABLE loop seeks to reduce source term uncertainty by studying experimentally the deposition and lift-off of Cs, I, Sr, and Ag under well-characterized conditions typical of a MGR primary circuit

  14. 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

  15. Experimental studies of flow induced vibrations of the fuel assembly for the PEC reactor

    International Nuclear Information System (INIS)

    The vibration behaviour of an assembly of seven mock-up fuel bundles of PEC reactor has been investigated. The assembly was excited by a parallel flow of water simulating sodium. The motion of the group (or of a single bundle in the group) has been measured in transverse sections detecting two orthogonal components of displacement. During the experiences the following parameters were varied: bundle foot and pads restraints, flow rate condition, coolant flow outlet conditions at the head of fuel bundles. Experimental data were processed in order to obtain: trajectories of three points of fuel bundle axis, power density spectra of measured vibration amplitudes, correlations between coolant flow rate and vibration amplitude R.M.S. (author)

  16. Design study of fuel circulating system for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    Design study of the fuel circulating system (FCS) for a tokamak experimental fusion reactor has been carried out. The FCS consists of main vacuum system, fuel gas refiners, isotope separators, neutral beam injector and fuel feed system. Cryopump vacuum system, fuel gas refiners of Pd-alloy membrane, and isotope separators of Pd-alloy membrane were mainly studied. Design parameters are: 16 cryosorption pumps, each-cryosorption-pump rate 3.3 x 105 l/sec, 12 fuel gas refiners, each-refiner effective surface area 820 cm2, 47 isotope separators (in 1st cascade) and 46 (in 2nd), each-separator effective surface area -- 2,600 cm2 (in 1st cascade) and -- 22,000 cm2 (in 2nd). (auth.)

  17. Probabilistic method for evaluation of reactivity margin of experimental very high temperature reactor

    International Nuclear Information System (INIS)

    A probabilistic method is proposed to evaluate in the core desigh stage the possibility that the safety criteria in reactivity margin are satisfied, taking into consideration the uncertainties in design calculation. In application of the method to design study cores of Experimental Very High Temperature Reactor, investigations are made for the relation between the design accuracy and the probability that the safety criteria in both shut-down and operation margins are satisfied. In conclusion, for the MARK-III core, with the correlation disregarded, the ratio of the standard deviations to the design values must be less than 0.79 and 5.3% respectively for the cold clean effective multiplication factor and the reactivity worths of control rods, burnable poisons and core temperature rise, in order that the probability is larger than 99.7% (three times the sigma limit). With the correlation regarded, the ratios must be considerably smaller. (author)

  18. 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)

  19. Determination of hydrazine in third loops of China experimental fast reactor by spectrophotometry

    International Nuclear Information System (INIS)

    The method for the determination of hydrazine by Uv-vis spectrophotometer was proposed. The coloration conditions and instrument parameters were also optimized. In HCl, hydrazine formed a yellow azine with para-dimethyl aminobenzaldehyde ((CH3)2NC6H4CHO), and then determined by spectrophotometer. The complex's maximum absorption was exhibited at 458 nm. The coloration effect was excellent in conditions of 1% HCl, 10 mL para-dimethyl aminobenzaldehyde and 10 minutes' developing time. A good liner relationship was obtained in the range of 5∼200 μg/L, and the recovery was (101.1±1.9)%. This method was used in the third loop of China experimental fast reactor with satisfactory results. (authors)

  20. Waste management for spent resin from Reactor Experimental Chileno number-sign 1 (RECH-1)

    International Nuclear Information System (INIS)

    A strategy is reported to find a waste form for temporary storage of spent resin arising from Reactor Experimental Chileno No. 1, RECH-1, according to radioactive waste management principles. As a first step, activity levels and contributions from long-lived fission products is obtained. The method that is developed is used to quantify the radioisotopes present in the resin to be allowed to follow them in the radioactive waste studies. As the activity is low, it is possible to find a way to dispose of large quantities of resin per drum. Two different processes were investigated: a mixtures technique and the double container immobilization method. Both were tested at laboratory and pilot plant scales. Results indicate that the double container method achieves great savings in economy of the management of spent resins

  1. Conceptual design of in-vessel components for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    This paper describes the results of KHI's design study on in-vessel components (i.s. first wall, tritium breeding blanket, radiation shielding, divertor and vacuum vessel) for the Tokamak type Fusion Experimental Reactor (FER) which is being developed by Japan Atomic Energy Research Institute. The mechanical configuration of the components and torus segmentation were determined, taking safety and reliability of the components and simplicity of remote-maintenance into consideration. Emphasis of the study was placed on developing a feasible first wall/blanket concept. Neutronic, electromagnetic and stress analysis were performed extensively in order to obtain necessary conditions to meet design criteria, and to adjust contradicting properties, e.g. tritium breeding performance, shell effect for plasma stabilization, structural integrity against various loads, and lifetime of the structure. Further, the fabrication process and joining technology for the first wall/blanket were examined and the most promising methods were selected. (author)

  2. Jules Horowitz Reactor Project- Fuel irradiation device, innovative instrumentation proposal for experimental phenomena real time measurement

    International Nuclear Information System (INIS)

    The fuel irradiation devices used for the tests or rods allow reproducing at small scales the conditions of the studied nuclear reactors (as LWR type). During the irradiation phase, the tested fuel rod can be stressed due to thermal, mechanical, nuclear effects which can modify its geometry (dilatation, swelling effects). After the test, the return to normal conditions can have as consequence the disappearance of the phenomenon or give access to partial information (final deformation). Generally, to follow the phenomena related to the irradiation phase, the experimental rod contained in the test device is instrumented with thermocouples and LVDT. As complement of this instrumentation, new sensors using innovating technologies are studied (deformation sensor integrating optical fibres). Through the example of a fuel irradiation device foreseen for the JHR, this paper aims to describe the present development of an innovating instrumentation with the objective to measure, in real time and under flux, the fuel rod deformation phenomena during a ramp test

  3. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. 2. Replacement of upper core structure

    International Nuclear Information System (INIS)

    In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test sub-assembly of MARICO-2 (material testing rig with temperature control) had bent onto the in-vessel storage rack as an obstacle and had damaged the upper core structure (UCS) in 2007. As a part of the restoration work, UCS replacement was begun at March 24, 2014 and was completed at December 17. In-vessel repair (including observation) for sodium-cooled fast reactors (SFRs) is distinct from that for light water reactors and necessitates independent development. Application of developed in-vessel repair techniques to operation and maintenance of SFRs enhanced their safety and integrity. There is little UCS replacement experience in the world and this experience and insights, which were accumulated in the replacement work of in-vessel large structure (UCS) used for more than 30 years, are expected to improve the in-vessel repair techniques in SFRs. (author)

  4. 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

  5. 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.

  6. 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

  7. 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 ...

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

    OpenAIRE

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

    2008-01-01

    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 ...

  9. 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

  10. Selected thermal and hydraulic experimentation in support of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    The ANS Reactor has unique thermal-hydraulic characteristics in comparison to other research and commercial reactors: Heavy water coolant, Parallel Rectangular channels (involute), Very small channel gap (1.27 mm), Very high velocity (25 m/s), Very high exit subcooling, Moderately high heat flux, High average power density. The objective was to determine experimentally the appropriate core thermal hydraulic limits at ANS conditions. Advanced Neutron Source (ANS) Thermal Hydraulic Test Loop (THTL) was designed to operate in 'Stiff', 'Soft' and 'Modified Stiff' Modes.Summary of Thermal Hydraulic Limit Testing and Analysis shows: FE data has been acquired at ANS typical flow velocities; An extensive OSV/OFI data base has been developed with a very broad parameter range, A modification of the Saha-Zuber correlation was proposed to account for reduced subcooling effects; Closeout activities include continued investigation of wider span test channels; Some testing for HFIR will be performed to evaluate the effect of reduced channel gap; Future plans called for additional testing at 3-core conditions, hot spot testing, etc. The Objective of Fuel Plate Stability Testing was to experimentally evaluate the structural response of ANS fuel plates to hydraulic loads. Summary of Fuel Plate Stability Testing shows: A Method Has Been Developed to Predict Structural Response of Fuel Plates to Hydraulic Loading Prediction of AP across plates Determine deflection/stress levels using structural analysis; ANS, Specific Conclusions are: no evidence of potential plate collapse in the coolant velocity range from 050 m/s, no evidence of plate flutter with coolant velocities below 33 m/s, local stress levels appear to dictate plate limits as opposed to plate deflection. The objective of Flow Blockage Testing was to experimentally determine local thermal and fluid. Summary of Flow Blockage Testing and Analysis showed: CFD code has been benchmarked against prototypic ANS flow conditions and

  11. Preliminary design and analysis of multi-functional fusion engineering experimental reactor plasma parameters based on regular Tokamak

    International Nuclear Information System (INIS)

    A multi-functional fusion test reactor concept named FDS-MFX (multi- functional engineering experimental reactor) proposed as a scenario option of China Fusion Engineering Test Reactor (CFETR) has been presented by FDS Team. FDS- MFX has been proposed for checking and validating the fusion DEMO reactor relevant technologies based on viable technologies. The preferred fusion core of FDS-MFX is regular Tokamak, with alternative choices such as spherical Tokamak and magnetic mirror, etc. In this paper, the core plasma parameters of FDS-MFX based on regular Tokamak were designed with the independently developed fusion system optimization and economic analysis code SYSCODE and analyzed based on the 'ITER Physics Basis'. We simulated the plasma equilibrium configuration and plasma discharge using the Tokamak Simulation Code (TSC); the result showed the core plasma parameters of FDS-MFX were preliminarily feasible. (authors)

  12. Evaluation of graphite/steam interactions for ITER [International Thermonuclear Experimental Reactor] accident scenarios

    International Nuclear Information System (INIS)

    This paper presents the results of an experimental/analytical study designed to determine the quantity of hydrogen generated during an accident involving coolant leakage into the plasma chamber of the International Thermonuclear Experimental Reactor (ITER). This hydrogen could represent a potential explosive hazard, provided the proper conditions exist, causing machine damage and release of radioactive material. We measured graphite/steam reaction rates for several graphites and carbon-based composites at temperatures between 1000 and 1700 degree C. The effects of steam flow rate and partial pressure were also examined. The measured reaction rates correlated well with two Arrhenius type relationships. We used the relationships for GraphNOL N3M in thermal model to determine that for ITER the quantity of hydrogen produced would range between 5 and 35 kg, depending upon how the graphite tiles are attached to the first wall. While 5 kg is not a significant concern, 35 kg presents an explosive hazard. 16 refs., 7 figs., 1 tab

  13. In vivo BNCT in experimental and spontaneous tumors at RA-1 reactor

    International Nuclear Information System (INIS)

    Within the search for new applications of Boron Neutron Capture Therapy (BNCT) and the basic research oriented towards the study of BNCT radiobiology to optimize its therapeutic gain, we previously proposed and validated the hamster cheek pouch oral cancer model and showed, for the first time, the success of BNCT to treat oral cancer in an experimental model. The staff of the Ra-1 Reactor (Constituyentes Atomic Center) adapted the thermal beam and physical set-up to perform in vivo BNCT of superficial tumors in small animals. We preformed a preliminary characterization of the thermal beam, performed beam only irradiation of normal and tumor bearing hamsters and in vivo BNCT of experimental oral squamous cell carcinomas in hamsters mediated by boron phenylalanine (BPA) and GB-10 (Na210B10H10). Having demonstrated the absence of radio toxic effects in healthy tissue and a therapeutic effect of in vivo BNCT in hamster cheek pouch tumors employing the Ra-1 thermal beam, we performed a feasibility study of the treatment by BNCT of 3 terminal cases of spontaneous head and neck squamous cell carcinoma in cats following the corresponding biodistribution studies. This was the first treatment of spontaneous tumors by BNCT in our country and the first treatment by BNCT in cats worldwide. This preclinical study in terminal cases showed significant tumor control by BNCT with no damage to normal tissue. (author)

  14. 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)

  15. Experimental thermal hydraulic facility for simulating LOCA behaviour of pressurised heavy water power reactor

    International Nuclear Information System (INIS)

    Experimental thermal hydraulic facility being set up adjacent to R and D Centre at Tarapur is a 13 MW full-elevation scaled down facility having the key components of PHT System of Pressurised Heavy Water Reactor (PHWR). The objective of the facility is to study thermal hydraulic behaviour of PHT System of PHWR by simulating various transients and accidental scenarios, to conduct safety related and operational transient studies and validation of various thermal hydraulic computer codes developed for analysis. The design of thermal hydraulic facility is based on the process parameters of a large PHWR with respect to fluid mass flux, transit time, flow velocity, pressure, temperature and enthalpy in PHT System. Experiments would be conducted in the facility to gain an improved understanding of the thermal hydraulic behaviour of large size PHWR during loss of coolant accident scenarios with forced and natural thermo-siphoning circulation modes etc. The data collected from the experiments would be used in validating computer codes developed for safety analysis. The facility is extensively instrumented to measure parameters such as temperature, pressure, flow, level, void-fraction at key locations. This paper gives the design philosophy used for scaling, design of major components of primary and secondary circuit of Experimental Thermal Hydraulic Facility and details of simulated experiments to be carried out. (author)

  16. Experimental investigation of thermal limits in parallel plate configuration for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m2, respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m2 and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45 degrees C) were maintained constant for these tests, while the loop was operated in a ''stiff''(constant flow) mode. Limited experiments were also conducted at 12 MW/m2 using a ''soft'' mode (near constant pressure-drop) for actual FE burnout tests and using a ''stiff' mode for true CHF tests, to compare with the original FE experiments

  17. Feasibility study of remote erosion measurement of first wall for Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    A remote measuring system has been developed for application to detection and dimensional measurement of eroded defects of protection tiles on the first wall in the Fusion Experimental Reactor. The present paper describes a remote defect measuring system, and results of the experimental study for the feasibility. The measuring system consists of a three-dimensional position sensor system and a manipulator to move it. The sensor system adopts a light sectional method using a laser slit light and an industrial TV camera, and an image fiber and a light guide are respectively used for transmission of an image and the light to protect the camera and the laser source against radiation. The system is able to detect defects of tiles in a large region and also to measure the form of only the defect part precisely. As a result of the experiment, it has been proved that the measuring system is practicable to the remote measurement of the first wall for FER. (orig.)

  18. 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)

  19. Measurement and evaluation of Corrosion Products deposition distribution in the Experimental Fast Reactor JOYO

    Energy Technology Data Exchange (ETDEWEB)

    Aoyama, Takafumi; Sumino, Kozo [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center; Masui, Tomohiko; Saikawa, Takuya

    1997-12-01

    The Corrosion Product (CP) is the major radiation source in the primary cooling system of an LMFBR plant. It is important to characterize and predict the CP behavior to reduce the personnel exposure dose due to CP deposition. The CP measurement was carried out in the Experimental Fast Reactor JOYO during the 11th annual inspection period when the accumulated reactor thermal power reached about 143 GWd. The CP deposition density was measured using a pure germanium detector. The plastic scintillation fiber (PSF) was applied for the gamma-ray dose rate distribution measurement and compared with the thermoluminescence dosimeter (TLD). The major results obtained by the CP measurements in JOYO are the follows: (1) The major CP nuclides deposited in the primary cooling system are {sup 54}Mn and {sup 60}Co. {sup 54}Mn is the dominant isotope and it tends to deposit in the cold leg region. On the other hand, {sup 60}Co deposits mainly in the hot leg region. The deposition density of {sup 54}Mn is about seven times as much as that of {sup 60}Co in the cold leg region and twice in the hot leg region. (2) The deposition densities of {sup 54}Mn and {sup 60}Co, and the gamma-dose rate were decreased from the last data in the previous annual inspection period mainly due to the short operation time and the longer cooling time. (3) The continuous gamma-ray dose rate distribution up to 10m can be measured by using the PSF in a few minutes. The PSF is suitable to measure the gamma-ray dose rate distribution in the maintenance work area where it is narrow and the mixture of gamma-ray sources from primary pipings and components. The data base of detailed gamma-ray dose rate distribution was greatly extended by the PSF. (author)

  20. 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.

  1. Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems: Revision 1

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnet systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

  2. 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.

  3. 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.

  4. Aspects of radiological protection for low power research reactor in the experimental teaching and learning

    International Nuclear Information System (INIS)

    It describes how learning by students and professional staff of nuclear facilities in the Czech Republic and Slovak Republic in the field of radiation protection within the reactor experiments at the school VR-1 Sparrow (CTU Faculty of Nuclear Sciences), where radiation protection is a natural essential part of every teaching and activities at the reactor. Reactor is presented and learning module 'Radiation Protection and Dosimetry' created for IAEA training courses - EERRI Group Fellowship Training Programme on Research Reactors (2010, 2011), which are intended for future training of personnel research reactors in IAEA Member States interested in the development of nuclear energy, but with little experience in this area. (authors)

  5. 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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Ali Khan, Liaquat; Ahmad, Nasir E-mail: epg.piaas@dgcc.org.pk; Zafar, M.S.; Ahmad, Ayaz

    2000-07-01

    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.

  8. 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

  9. 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

  10. 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...

  11. 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)

  12. 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.)

  13. 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.

  14. Experimental investigations of steam bubble entrainment in the down-flow cavity of a natural circulation heat only reactor

    International Nuclear Information System (INIS)

    The ZHM-2 (Zittauer Heizreaktormodell, Zittau Heat Only Reactor Model) experimental facility is described. This facility is used for studying entrainment of gas bubbles during 1800 flow inversion from the up-flow into the down-flow cavity of a natural circulation heat only reactor. Volume gas content, bubble speed, and bubble size can be determined using a special conductivity probe. The gas portion entrained in the down-flow cavity PSI is presented as a function of water level, gas content distribution in the up-flow cavity, and geometry of the upper inverting region. (author)

  15. Experimental testing of reduced-scale seismic isolation bearings for the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    A series of tests of reduced-scale seismic isolation bearings undertaken in support of the development of a seismic isolation concept for the Advanced Liquid Metal Reactor (ALMR) is described. A procurement specification applicable to both full-size and reduced-scale bearings was developed by the program participants and used to purchase bearings of four different designs from two manufacturers. The high-damping rubber isolators were subjected to horizontal, vertical, and failure tests designed to quantify their mechanical properties both within the range of design loads and displacements as well as to establish their margins before failure. The test results show that bearings from both manufacturers provide stable and repeatable behavior with minor variations in stiffness and damping as a function of loading frequency and load history. None of the bearings showed substantial variation in properties due to changes in axial load. All of the bearings exhibited exceptional behavior when loaded beyond the design level, with displacement margins greater than 3 and force margins greater than 4. This test program provides a thorough data-set for further analytical and experimental validations of the seismic isolation concept for the ALMR. (author)

  16. 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)

  17. 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)

  18. 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)

  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. 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.

  1. 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

  2. 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

  3. 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.

  4. Comparison of Heavy Water Reactor Thermalhydraulic Code Predictions with Small Break LOCA Experimental Data

    International Nuclear Information System (INIS)

    Activities within the frame of the IAEA's Technical Working Group on Advanced Technologies for HWRs (TWG-HWR) are conducted in a project within the IAEA's subprogramme on nuclear power reactor technology development. The objective of the activities on HWRs is to foster, within the frame of the TWG-HWR, information exchange and cooperative research on technology development for current and future HWRs, with an emphasis on safety, economics and fuel resource sustainability. One of the activities recommended by the TWG-HWR was an international standard problem exercise entitled Intercomparison and Validation of Computer Codes for Thermalhydraulics Safety Analyses. Intercomparison and validation of computer codes used in different countries for thermalhydraulics safety analyses will enhance the confidence in the predictions made by these codes. However, the intercomparison and validation exercise needs a set of reliable experimental data. Two RD-14M small break loss of coolant accident (SBLOCA) tests, simulating HWR LOCA behaviour, conducted by Atomic Energy of Canada Ltd (AECL), were selected for this validation project. This report provides a comparison of the results obtained from eight participating organizations from six countries (Argentina, Canada, China, India, Republic of Korea, and Romania), utilizing four different computer codes (ATMIKA, CATHENA, MARS-KS, and RELAP5). General conclusions are reached and recommendations made.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-12-01

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

  6. Design study of superconducting magnets for tokamak experimental fusion reactor, (1)

    International Nuclear Information System (INIS)

    Design study has been made of superconducting magnets for a Tokamak experimental fusion reactor: toroidal field magnet design, poloidal field magnet design, refrigeration system design, magnet safety analysis, and magnet assembling and disassembling system design. A maximum toroidal field in the coil is 11.0 T, providing 5.5 T at plasma center. Nb3Sn superconducting cable is used to attain the toroidal field of 11 T. The coil bore is 7.3 x 11.2 m, and the coil shape is deformed constant-tension D-shape. The magnetomotive force is 185.6 MAT, and the operational current is 25.9 kA. In poloidal field magnet design, the coil is pancake-wound Nb3Sn conductor. The conductor is enclosed in Ti-alloy sheath, which serves also as helium containment vessel. The conductor is cooled by forced flow supercritical helium of 7 atm and 4.6 K, and the operational current is 25 -- 27 kA. (author)

  7. 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.

  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. 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

  10. 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

  11. 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.)

  12. 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.)

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

    Science.gov (United States)

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

    2003-01-01

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

  14. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors

    International Nuclear Information System (INIS)

    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)

  15. Experimental, kinetic and numerical modeling of hydrogen production by catalytic reforming of crude ethanol over a commercial catalyst in packed bed tubular reactor and packed bed membrane reactor

    International Nuclear Information System (INIS)

    The demand for hydrogen energy has increased tremendously in recent years essentially because of the increase in the word energy consumption as well as recent developments in fuel cell technologies. The energy information administration has projected that world energy consumption will increase by 59% over the next two decades, from 1999 to 2020, in which the largest share is still dominated by fossil fuels (oil, natural gas and coal). Carbon dioxide (CO2) emissions resulting from the combustion of these fossil fuels currently are estimated to account for three-fourth of human-caused CO2 emissions worldwide. Greenhouse gas emission, including CO2, should be limited, as recommended at the Kyoto Conference, Japan, in December 1997. In this regard, hydrogen (H2) has a significant future potential as an alternative fuel that can solve the problems of CO2 emissions as well as the emissions of other air contaminants. One of the techniques to produce hydrogen is by reforming of hydrocarbons or biomass. Crude ethanol (a form of biomass, which essentially is fermentation broth) is easy to produce, is free of sulphur, has low toxicity, and is also safe to handle, transport and store. In addition, crude ethanol consists of oxygenated hydrocarbons, such as ethanol, lactic acid, glycerol, and maltose. These oxygenated hydrocarbons can be reformed completely to H2 and CO2, the latter of which could be separated from H2 by membrane technology. This provides for CO2 capture for eventual storage or destruction. In the case of using crude ethanol, this will result in negative CO2, emissions. In this paper, we conducted experimental work on production of hydrogen by the catalytic reforming of crude ethanol over a commercial promoted Ni-based catalyst in a packed bed tubular reactor as well as a packed bed membrane reactor. As well, a rigorous numerical model was developed to simulate this process in both the catalytic packed bed tubular reactor and packed bed membrane reactor. The

  16. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

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

    OpenAIRE

    Alessandro Alemberti; Fabio Berra; Davor Grgic; Graydon Yoder; Stefano Monti; Paride Meloni; Davide Papini; Fosco Bianchi; Marco Ricotti; Roberta Ferri; Cinzia Congiu; Gustavo Cattadori; Andrea Achilli; Bojan Petrovic; Andrea Maioli

    2009-01-01

    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 bu...

  18. A small-scale experimental reactor combined with a simulator for training purposes

    International Nuclear Information System (INIS)

    The authors discuss how a small-scale reactor combined to a training simulator can be a valuable aid in all forms of training. They describe the CEN-based SILOETTE reactor in Grenoble and its combined simulator. They also take a look at prospects for the future of the system in the light of experience acquired with the ARIANE reactor and the trends for the development of simulators for training purposes

  19. Experimental study on biomass gasification in a double air stage downdraft reactor

    International Nuclear Information System (INIS)

    This work presents an experimental study of the gasification of a wood biomass in a moving bed downdraft reactor with two-air supply stages. This configuration is considered as primary method to improve the quality of the producer gas, regarding its tar reduction. By varying the air flow fed to the gasifier and the distribution of gasification air between stages (AR), being the controllable and measurable variables for this type of gasifiers, measuring the CO, CH4 and H2 gas concentrations and through a mass and energy balance, the gas yield and its power, the cold efficiency of the process and the equivalence ratio (ER), as well as other performance variables were calculated. The gasifier produces a combustible gas with a CO, CH4 and H2 concentrations of 19.04, 0.89 and 16.78% v respectively, at a total flow of air of 20 Nm3 h-1 and an AR of 80%. For these conditions, the low heating value of the gas was 4539 kJ Nm-3. Results from the calculation model show a useful gas power and cold efficiency around 40 kW and 68%, respectively. The resulting ER under the referred operation condition is around 0.40. The results suggested a considerable effect of the secondary stage over the reduction of the CH4 concentration which is associated with the decreases of the tar content in the produced gas. Under these conditions the biomass devolatilization in the pyrolysis zone gives much lighter compounds which are more easily cracked when the gas stream passes through the combustion zone. -- Highlights: → Obtained results an important for a better phenomenological understanding of processes occurring in two stage gasification reactors. → The air flow is the fundamental parameter in the operation of downdraft gasifiers. → CH4 reduction is associated with a decreases in the tar content. → An enhancement in the thermal cracking of tar is carried out in the two-air downdraft gasifier.

  20. 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

  1. 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.

  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. A Design Study on Experimental Power Reactor Core Fueled with UO2 CFP

    International Nuclear Information System (INIS)

    A neutronic study on core design of a 300 MWt EPR was performed. In this study the use of 4.8% enriched UO2 coated fuel particle was analyzed. The design was then compared to 5% enriched UO2 pin fueled EPR based on existing PWRs. Both reactors are operated with single batch refueling system with a cycle length of 3 years. The core physics parameters analyzed were : effective multiplication factor in a cycle, flux distributions and cycle burnup. The results of calculation showed that the core effective multiplication factor for reactor with fuel compact can be maintained at 1.2841 at beginning of cycle (BOC) and 1.0060 at end of cycle (EOC). As for the UO2 pin fueled reactor, the effective multiplication factor was 1.1927 at BOC and 1.0514 at EOC. The size of active core for the CFP fueled reactor were 320 cm in height and 320 cm in diameter. As for pin fueled reactor, the height was 200 cm and diameter was 180 cm. The results of calculations showed that neutron flux distribution was quite flat for both types of reactor designs, although the volume of CFP fueled reactor was 5 times as big as the pin fueled reactor

  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. 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.

  6. 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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'JOYO'. Recovery of MARICO-2 sample part

    International Nuclear Information System (INIS)

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. The following items are introduced here: (1) summary of restoration work and overall process of restoration work, (2) recovery operation of MARICO-2 sample part, (3) exchange of the upper core structure that was conducted this year, and (4) results of recovery of MARIKO-2 sample part. (A.O.)

  12. Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 1. MARICO-2 subassembly retrieval work

    International Nuclear Information System (INIS)

    At Joyo reactor MK-III core in May 2007, due to the design deficiencies of the disconnect mechanism of the holding part and the sample part of the experimental apparatus with instrumentation lines (MARICO-2), a disconnect failure incident occurred in the sample part after irradiation test. The deformation of the sample part due to this failure incurred its interference with the lower surface of reactor core upper structure and the holddown axis body. By this, the operating range of the rotary plug was restricted, leading to the partial inhibition of the fuel exchange function that precluded the access to 1/4 of the assemblies of the reactor core. In face of restoration work, the preparation for restoration such the exchange of upper core structure, and the recovery of MARICO-2 sample part are under way. This paper introduces the progress of restoration work and the future work plan, with a focus on the outline of overall restoration work, the method / problems / measures for MARICO-2 sample part recovery operations, and fabrication of sample part recovery device. (A.O.)

  13. Experimental analysis of oxygen-methane combustion inside a gas turbine reactor under various operating conditions

    International Nuclear Information System (INIS)

    The oxygen-methane diffusion flame taking place in a gas turbine reactor was investigated experimentally with emphasis on flame stability. The oxidizer is a mixture of O2 and CO2 and the oxy-combustion process was studied at different equivalence ratios ranging from Φ = 0.5 to 1.0 and different O2/CO2 mixture composition (100/0, 80/20, 60/40, 50/50, 40/60, 30/70 and 25/75). The flame blowout condition was achieved through the reduction of oxygen percentage in the oxidizer mixture. Measurements were obtained for the flue gas temperature and concentration as well as flame visualization. It was found that the flame is very stable at the equivalence ratio of 0.65. At this ratio, the flame blows out at an O2/CO2 blending ratio of 22/78 for the case of fuel flow rate of 6 L/min and at a blending ratio of 21/79 for the case of fuel flow rate of 9 L/min. Attempts for operating the burner with less than 21% O2 were unsuccessful at all ranges of the operating parameters and resulted in unstable operation and blowout. Moreover, it was observed that the stabilization behavior did not change significantly with the variation of the fuel volume flow rate. It was also found that both flame and flue gas temperatures are reduced with the increase of the equivalence ratio. - Highlights: • Stability and characteristics of oxy-combustion diffusion flame were investigated. • The flame blowout conditions was determined. • Visualization of flame is carried out and the exhaust gas temperature is measured. • It was found that the most stable flame is at an equivalence ratio of 0.65. • Attempts for operating the burner with less than 21% O2 were unsuccessful

  14. Experimental analysis of a combustion reactor under co-firing coal with biomass

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Fabyo Luiz; Bazzo, Edson; Oliveira Junior, Amir Antonio Martins de [Universidade Federal de Santa Catarina, Florianopolis, SC (Brazil). LabCET], e-mail: ebazzo@emc.ufsc.br; Bzuneck, Marcelo [Tractebel Energia S.A., Complexo Termeletrico Jorge Lacerda, Capivari de Baixo, SC (Brazil)], e-mail: marcelob@tractebelenergia.com.br

    2010-07-01

    Mitigation of greenhouse gases emission is one of the most important issues in energy engineering. Biomass is a potential renewable source but with limited use in large scale energy production because of the relative smaller availability as compared to fossil fuels, mainly to coal. Besides, the costs concerning transportation must be well analysed to determine its economic viability. An alternative for the use of biomass as a primary source of energy is the co-firing, that is the possibility of using two or more types of fuels combined in the combustion process. Biomass can be co-fired with coal in a fraction between 10 to 25% in mass basis (or 4 to 10% in heat-input basis) without seriously impacting the heat release characteristics of most boilers. Another advantage of cofiring, besides the significant reductions in fossil CO{sub 2} emissions, is the reduced emissions of NO{sub x} and SO{sub x}. As a result, co-firing is becoming attractive for power companies worldwide. This paper presents results of some experimental analysis on co-firing coal with rice straw in a combustion reactor. The influence of biomass thermal share in ash composition is also discussed, showing that alkali and earth alkaline compounds play the most important role on the fouling and slagging behavior when co-firing. Some fusibility correlations that can assist in the elucidation of these behavior are presented and discussed, and then applied to the present study. Results show that for a biomass thermal share up to 20%, significant changes are not expected in fouling and slagging behavior of ash. (author)

  15. 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.)

  16. 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.)

  17. 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)

  18. 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...

  19. Stand complex of experimental investigations of the NPP safety and perspective directions of nuc,ear reactor development

    International Nuclear Information System (INIS)

    A stand complex for experimental investigations of the safety of NPP with the WWER type reactors is described. The complex includes: the Parameter-M stand for studying the behavior of fuel elements and cores in emergency situations; the Lava-P stand for studying the steam explosion during interaction of corium with water; the MRT-1 material-science stand. The main performances of these stands and first obtained results are presented

  20. 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

  1. 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...

  2. Interpretation of TRIGA reactor experimental data with SIMMER-III code for RELAP5 model evaluation and transient analysis

    International Nuclear Information System (INIS)

    The experimental data obtained in a test campaign carried out in the TRIGA reactor of ENEA/Casaccia are being used to validate a coupled RELAP5/PARCS numerical model for the thermal-hydraulic and dynamic simulation of the reactor. The tests conducted at different power levels provided the temperature measurements at the core inlet and outlet that allow the evaluation of DT through the core at different radial positions. In order to interpret the experimental data for the evaluation of total water mass flow rate through the core in natural circulation, several calculations have been performed with the SIMMER-III code (CFD two-dimensional code) at different core power levels trying to reproduce the experimental measurements. The results of SIMMER-III code were used to fit and validate the simplified 1-D model of the RELAP5 code used for thermal-hydraulic transient analysis of TRIGA reactor. Finally, this paper presents the interpretation of some reactivity transients using this improved T/H model with the RELAP5 point-kinetics neutronic model. (author)

  3. Tests of experimental fuel elements by the method of nuclear-thermal pulse loadings in 'HYDRA' reactor

    International Nuclear Information System (INIS)

    The results of tests of experimental fuel elements with uranium dioxide fuel composition embedded in Al and Zr matrix with the enrichment from 90% to 36% in respect to U-235 performed at the pulse 'HYDRA' reactor are presented in this paper. Testing is performed in the frame-work of extensive research program studying the behavior of fuel elements (FE) of research and mini nuclear power systems in case of practically immediate energy release in the fuel taking place during the RIA-type accidents. Duration of the neutron pulse when testing in 'HYDRA' reactor is from 7 to 20 ms. The methods of diagnostics of the state of FE prior to and after testing in the reactor are developed and verified. Mathematical model describing temperature fields inside the FE in the process of testing. and accounting for non-uniformity of fuel composition has been developed in order to summarize experimental results. Experimental data on the limiting values of the energy density leading to deformation and degradation of FE depending on the type of fuel composition have been obtained and the mechanisms for the development of these processes have been determined. The nature of physical-chemical processes taking place in the fuel composition and fuel cladding depending on material composition under different levels of energy deposition is demonstrated. The data on hydrogen generation and radioactive product release out of fuel after failure of FE are presented. (author)

  4. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  5. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

  6. Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

    International Nuclear Information System (INIS)

    Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

  7. 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...

  8. Determination of the theoretical and experimental zero-power frequency response of Ghana Research Reactor-1

    International Nuclear Information System (INIS)

    The frequency response measurements of a reactor at low power help in determining the kinetic parameters of a reactor and ultimately in investigating its stability with respect to small perturbations in reactivity. In this report, we present the results of the zero-power frequency response measurements of GHARR-1 by rod method and its analytical analogue. The comparison in calculated and measured values is reasonably good in the frequency range used (author)

  9. 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.

  10. 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.

  11. Experimental and theoretical vibration analysis by noise measurements of a sodium cooled reactor

    International Nuclear Information System (INIS)

    Studies of mechanical vibrations of primary components and their correlations with pressure and neutron flux at the compact sodium cooled reactor KNK (20 MWe), Karlsruhe, gave substantial information of the movement of the reactor vessel and its internals. Displacement measurements and noise measurements of pressure and neutron flux were performed in the power range of 27%-100%. The maximum of the power spectral densities together with the observed natural frequencies of reactor components were investigated. By calculating the power spectral densities with different filter bandwidths deterministic and stochastic parts of the spectra were separated. The most powerful peaks in the spectra were flow-induced stochastic spectra parts as well as two deterministic lines, excited mechanically by the pumps. Some of these maxima could also be observed by correlating them with the noise of pressure and neutron flux. By means of a model the influence of the process dynamics on the power density spectra of the measured variables was checked. The comparison of power spectral densities and coherences of mechanical vibrations of the vessel at different power levels showed natural frequencies down to 25 cps. Phases of cross power spectral densities revealed a combined horizontal and vertical movement of the reactor vessel. The impact of a free falling control rod was successfully used as a mechanical excitation of the system. The measurements showed natural frequencies of various reactor internals (e.g. core support, shock shield). Tests with variable pump speed gave information on global mechanical characteristics of the reactor system. (Auth.)

  12. 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

  13. 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.

  14. Experimental verification of methods and codes used in design studies of new reactor concepts and improved in-core fuel management schemes

    International Nuclear Information System (INIS)

    The paper gives a short description of the zero power heavy water research reactor RB at the Nuclear Engineering Department of the Boris Kidric Institute of Nuclear Sciences in Vinca. It describes shortly different experiments and different modifications performed on this facility during the last two decades. In particular, it concentrates on the possibility of using this reactor to verify experimentally methods and codes used in design studies new reactor concepts and improved in-core fuel management schemes. When calculating axially inhomogeneous reactor core, for example, procedures for determining reactor lattice cell parameters and overall reactor parameters can be combined in different ways. Since it was possible to simulate such an axially inhomogeneous core on the RB reactor, experimental verification of different calculational models has been performed, Suggestions, given as the conclusion of these experiments, refer not only to the situation when axial inhomogeneity originates from different initial composition of a reactor core, but also to the most common situation when the axial inhomogeneity is caused oy nonuniform depletion during reactor operation. Finally, the paper points out a number of other interesting and useful experiments which can be performed on the zero power reactor RB and indicates its potential use in international cooperative programs. (author)

  15. Experimental and theoretical vibration analysis by noise measurements of a sodium cooled reactor

    International Nuclear Information System (INIS)

    Studies of mechanical vibrations of primary components and their correlations with pressure and neutron flux at the compact sodium cooled reactor KNK (20 MWe), Karlsruhe, gave substantial information on the movements of the reactor vessel and its internals. The maximum of the power spectral density together with the observed natural frequencies of reactor components were investigated. The most powerful peaks in the spectra were flow-induced stochastic spectra parts as well as two deterministic lines, excited mechanically by the pumps. Some of these maxima could also be observed by correlating them with the noise of pressure and neutron flux. By means of a model the influence of the process dynamics on the power density spectra of the measured variables was checked. The comparison of power spectral densities and coherences of mechanical vibrations of the vessel at different power levels showed natural frequencies down to 25 cps. Phases of cross power spectral densities revealed a combined horizontal and vertical movement of the reactor vessel. The impact of a free falling control rod was successfully used as a mechanical excitation of the system. The measurements showed natural frequencies of various reactor internals (e.g. core support, shock shield). Tests with variable pump speed gave information on global mechanical characteristics of the reactor system. The use of Donnell's equation for the inner and outer reactor vessel and the shock shield showed good agreement between theoretical and measured natural frequencies and explained the complex structure of the spectra. The employed measurement principles are being investigated with regard to their application for a vibration surveillance system

  16. Experience of primary cooling system modification to increase heat removal capability in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    The purpose of the MK-III program is to upgrade the irradiation capability of the JOYO experimental sodium-cooled fast reactor. As a result, the neutron flux density of the core was increased and the reactor thermal power was increased to 140 MWt from the originally designed 100 MWt. To accommodate the increased thermal power, the flow rates of sodium coolant in the primary and secondary systems were increased by 20 % and 10 %, respectively. Also, all intermediate heat exchangers and dump heat exchangers were replaced with new ones. During this replacement, molten sodium and fuel were retained in the reactor vessel. Consequently, the primary cooling system and cover gas boundaries had to be maintained to prevent impurity ingress to the sodium system. During the replacement, the seal bag method, impurity concentration monitoring of cover gas, and low-pressure control of cover gas were applied to prevent damage to existing components and systems. A roller cutter was used for cutting large diameter pipes to prevent ingress of cuttings. The measures taken to reduce the radiation exposure were a lowering of the surrounding dose rate through the use of temporary shielding, shortening of the operation time near the high dose rate area by first doing thorough training, and the employment of protection equipment to avoid contamination. The replacement of components was completed without major trouble, and methods applied for the replacement proved to be effective in the operation and maintenance of sodium cooled reactors. (author)

  17. 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

  18. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-08-15

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

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

    DEFF Research Database (Denmark)

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

    2004-01-01

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

  1. 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

  2. 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

  3. TIT reactor laboratory course using JAERI and PNC large experimental facilities

    International Nuclear Information System (INIS)

    This report is presented on a reactor laboratory course for graduate students using large facilities in national laboratories in Japan. A reactor laboratory course is offered every summer since 1990 for all graduate students in the Nuclear Engineering Course in Tokyo Institute of Technology (TIT), where the students can choose one of the experiments prepared at Japan Atomic Energy Research Institute (JAERI), Power Reactor and Nuclear Fuel Development Corporation (PNC) and Research Reactor Institute, Kyoto University (KUR). Both JAERI and PNC belong to Science and Technology Agency (STA). This is the first university curriculum of nuclear engineering using the facilities owned by the STA laboratories. This type of collaboration is promoted in the new Long-Term Program for Research, Development and Utilization of Nuclear Energy adopted by Atomic Energy Commission. Most students taking this course reported that they could learn so much about reactor physics and engineering in this course and the experiment done in large laboratory was a very good experience for them. (author)

  4. High pressure reactor water loop for the experimental studies in the field of water chemistry and corrosion

    International Nuclear Information System (INIS)

    At the Nuclear Research Institute in Rez, an experimental reactor pressurized-water loop was built up on basis of a plan of technical development of nuclear power stations with PWR's. This loop is intended for further research and control of conditions in reactor circuits exposed to radiation. The loop (RVS-3) comprises a closed pressurized-water circuit with forced circulation and has a test section located in the core of the VVR-S type reactor, and filtration and measuring circuits. The loop is so equipped that it is possible to operate the loop with the following nominal parameters: working overpressure 15.7 MPa; working temperature 350 deg. C; working medium chemically treated water; flowrate 10,000 kg/hr. The paper summarizes some results of experimental work carried out in NRI during the recent period in the field of electromagnetic filter development, technology of boric acid control and remote electro-chemical methods of measurement of instantaneous corrosion rate of materials. The results of the electromagnetic filter efficiency testing and the development work on the boric acid control by the thermal regeneration of strong base anion resins are presented. The results of the preliminary tests of corrosion of fuel cladding material (zirconium alloy Zr-Nb) in an out-of-pile loop with parameters corresponding to PWR primary circuit is also included. (author)

  5. 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)

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

    International Nuclear Information System (INIS)

    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.

  7. 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.

  8. 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

  9. 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)

  10. 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

  11. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  13. 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.

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

    International Nuclear Information System (INIS)

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

  15. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  16. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    Science.gov (United States)

    Wang, Shijia; Wang, Shaojie

    2015-04-01

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ˜30 % , with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by 60 % , with the central pressure also significantly raised.

  17. Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

    International Nuclear Information System (INIS)

    The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITER inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ∼30%, with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by  60%, with the central pressure also significantly raised

  18. 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

  19. 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…

  20. 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)

  1. 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

  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 safety...

  3. 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)

  4. 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.

  5. 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)

  6. 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

  7. 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)

  8. Experimental stability maps for a two-phase natural circulation reactor with and without void-reactivity feedback effect

    International Nuclear Information System (INIS)

    Highlights: • A scaled experimental facility is designed based on the sound scaling approach. • Experiments are performed to identify the instability phenomena. • Four heater rods are used to simulate the chaotic flashing phenomena. • Experiments are performed with and without void-reactivity feedback. • Stability maps are obtained at different system pressures. -- Abstract: The new small-scaled light water reactor design, known as integral modular water reactor (IMR), is susceptible to flow instabilities due to two-phase natural circulation inside the reactor pressure vessel (RPV). Flow instabilities may be amplified due to strong interaction between the flow and the core power through the void-reactivity feedback mechanism. During the start-up of the IMR, system pressure is low. At low pressure, the density ratio is quite high, which leads to large variation in void fraction due to change in the flow quality. In the IMR design, the long riser and large volume of water can lead to thermal non-equilibrium between the phases due to the significant variation of the saturation temperature along flow direction. This can result in flow oscillations at certain operating conditions. In order to understand and identify the instability phenomena during the start-up of the reactor, a scaled experimental facility is designed based on the sound scaling approach. The scaling laws are used to obtain design parameters to maintain the similarities between the prototype and the experimental facility. Four heater rods are used to simulate the chaotic flashing phenomena. The steady state tests are performed with and without void-reactivity feedback at different system pressures. The flow is stable below a certain core power regardless of the channel inlet subcooling. The region of stability grows in size as the core power is increased. The unstable region reduces significantly at high pressure compared to low pressure case. A new approach is presented to simulate the void

  9. 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

  10. Basic experimental stimulation of a traveling wave direct energy converter for a D-3He fusion reactor

    International Nuclear Information System (INIS)

    We simulated a traveling wave direct energy converter for application in a D-3He fusion reactor. The deceleration of a simulated ion beam was studied, and the dependence of the efficiency of the decelerator on its length was examined. The effectiveness of spatial variation in the phase velocity of the traveling wave was demonstrated experimentally. Optimization of the structure of the decelerator is discussed with the aid of numerical simulations. By employing the proper decelerator structure, the efficiency per unit length of about 0.2 with the normalized decelerator length of about 2.5 is expected. (author)

  11. Experimental and analytical studies of a deeply embedded reactor building model considering soil-building interaction. Pt. 1

    International Nuclear Information System (INIS)

    The purpose of this paper is to describe the dynamic characteristics of a deeply embedded reactor building model derived from experimental and analytical studies which considers soil-building interaction behaviour. The model building is made of reinforced concrete. It has two stories above ground level and a basement, resting on sandy gravel layer at a depth of 3 meters. The backfill around the building was made to ground level. The model building is simplified and reduced to about one-fifteenth (1/15) of the prototype. It has bearing wall system for the basement and the first story, and frame system for the second. (orig.)

  12. 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

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

    OpenAIRE

    DANICA BRZIC; DESISLAVA AHCHIEVA; MIRKO PEGLOW; STEFAN HEINRICH

    2007-01-01

    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 sele...

  14. Conceptual design of Fusion Experimental Reactor (FER) based on an advanced scenario of plasma operation and control

    International Nuclear Information System (INIS)

    The Fusion Experimental Reactor (FER) which is being developed at JAERI as a next-generation tokamak following JT-60 has the major purpose of realizing a self-ignited, long-burning DT plasma and demonstrating engineering feasibility. The paper emphasizes the advanced scenario of FER plasma operation and control and the advantage in engineering design made possible by the scenario. The FER concept is discussed, which is based on quasi-steady-state operation by a lower-hybrid-wave current drive or steady-state operation by three candidate radiofrequency waves, impurity control by a cold and dense divertor plasma and vertical position control of a highly elongated plasma. (author)

  15. 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.

  16. Reference core design Mark-III of the experimental multi-purpose, high-temperature, gas-cooled reactor

    International Nuclear Information System (INIS)

    The reactivity control system is one of the important items in reactor design, but it is much restricted by structural design of fuel element and pressure vessel in the experimental multi-purpose, high-temperature reactor. Preceding the first conceptual design of the reactor, therefore, the reactivity control system composed of control rod, burnable poison and reserve shutdown system in Mark-II design was re-studied, and several improvements were indicated. (1) The diameter of control rods must be as large as possible because it is impossible to increase the number of control rods. (2) The accuracy in estimation of the reactivity to be compensated with control rods is important because of the mutual interference of pair control rods with the twin configuration in a fuel element. (3) The improvement of core performance in burnup is accompanied by the reduction of design margin for control rods. (4) Increase of the reactivity to be compensated with the burnable poison leads to increase of the core reactivity recovery with burnup, and the assertion of the decrease for recovery of reactivity leads to increase of the temperature dependency of reactivity compensated with control rods. (5) Reduction of reactivity to be compensated with control rods is thus limited by cancellation of the effects in the reactivity recovery and the reactivity temperature dependency. (6) The reserve shutdown system can be designed with margin under the condition of excluding the reactivity of burnup from that to be compensated. (auth.)

  17. 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)

  18. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    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)

  20. 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