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Sample records for experimental reactor project

  1. The 'Reacteur Jules Horowitz': a new experimental reactor project

    International Nuclear Information System (INIS)

    Frachet, S.; Ballagny, A.

    1999-01-01

    The Jules Horowitz Reactor (RJH) is a new research reactor project dedicated to materials and nuclear fuel testing, the location of which is foreseen at the CEA-CADARACHE site, and the start-up in 2006. The launching of this project originated from a double finding: The development of nuclear power plants aimed at satisfying the energy needs of the next century, cannot be envisaged without experimental reactors which are unrivaled for the validation of new concepts of nuclear fuels, materials, and components as well as for their qualification under irradiation. The existing experimental reactors are 30 to 40 years old and it is advisable to examine henceforth the necessity for and the nature of a new reactor to take over and replace, at the beginning of next century, the reactors shut-down in the mean time or at the very end of their lives. Within this framework, the CEA has undertaken, in the last years, a study on the mid and long term irradiation needs, to determine the main features and performances of this new reactor. The concept of the reactor will have to fulfill the thermal neutron irradiation requirements as well as the fast neutron experimental needs, with a great potential versatility for any new irradiation programs. The reactor project selected among several different concepts, is finally a light water pool concept, with 100 MW thermal power. It could reach neutronic fluxes twice those of present French reactors, and allows for many irradiations in and around the core, under high neutron fluxes. The reactor will satisfy the highest level of safety in full accordance with international safety recommendations and the French safety approach for this kind of nuclear facility, thus giving an added safety margin keeping in mind the versatility of research reactors. The feasibility studies have been focused on the following most important items: neutronic and thermalhydraulic studies on alternative core designs, with or without added pressurization

  2. Industrial opportunities on the International Thermonuclear Experimental Reactor (ITER) project

    International Nuclear Information System (INIS)

    Ellis, W.R.

    1996-01-01

    Industry has been a long-term contributor to the magnetic fusion program, playing a variety of important roles over the years. Manufacturing firms, engineering-construction companies, and the electric utility industry should all be regarded as legitimate stakeholders in the fusion energy program. In a program focused primarily on energy production, industry's future roles should follow in a natural way, leading to the commercialization of the technology. In a program focused primarily on science and technology, industry's roles, in the near term, should be, in addition to operating existing research facilities, largely devoted to providing industrial support to the International Thermonuclear Experimental Reactor (ITER) Project. Industrial opportunities on the ITER Project will be guided by the amount of funding available to magnetic fusion generally, since ITER is funded as a component of that program. The ITER Project can conveniently be discussed in terms of its phases, namely, the present Engineering Design Activities (EDA) phase, and the future (as yet not approved) construction phase. 2 refs., 3 tabs

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

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1986-05-01

    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 10 15 n/cm 2 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) [pt

  4. 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)

    Pavicevic, M.

    1963-07-01

    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

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

    International Nuclear Information System (INIS)

    1975-01-01

    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

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

    International Nuclear Information System (INIS)

    Boing, L.E.

    1989-12-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.

    1989-12-01

    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.

  8. Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

    International Nuclear Information System (INIS)

    Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

    1994-01-01

    ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

  9. Organization of the ITER [International Thermonuclear Experimental Reactor] Project - Sharing of information and procurements

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1990-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is expected to fully confirm the scientific feasibility and to address the technological feasibility of fusion power. Consequently, the machine must be designed for controlled ignition and extended burn of deuterium-tritium plasma. It must also demonstrate and perform integrated testing of components required to utilize fusion power for practical purposes. Cooperation among four countries/organizations (United States, Soviet Union, Japan, and EURATOM) to build a single experimental reactor will reduce the cost for each country and provide an international pool of scientific and engineering resources. This paper describes ITER organization for conceptual design activity, schedule for conceptual design activities, ITER operating parameters, conceptual project schedule and cost, future plans, basic principles and problems related to task sharing, and basic principles in handling of intellectual property

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

    International Nuclear Information System (INIS)

    Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

    2002-01-01

    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 Li 2 TiO 3 , 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 U MP'. Scientific and technical support of

  11. ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

    International Nuclear Information System (INIS)

    Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

    2008-01-01

    In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

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

    International Nuclear Information System (INIS)

    1993-01-01

    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

  13. Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Boing, L.E.; Aldana, J.

    1997-03-01

    The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

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

    International Nuclear Information System (INIS)

    Lahodova, Zdena; Viererbl, Ladislav; Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  16. The international thermonuclear reactor project

    International Nuclear Information System (INIS)

    James, T.R.

    1993-01-01

    The International Thermonuclear Experimental Reactor Project is a 6-year collaborative effort involving the U.S., Europe, Japan, and the Russian Federation to produce a detailed engineering design for the next-step fusion device

  17. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca

    International Nuclear Information System (INIS)

    Novakovic, M.

    1961-12-01

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO 2 with beryllium cladding, cooled by CO 2 under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO 2 . This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. 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)

    Ishiguro, Y.

    1983-01-01

    In order to assure the continued utilization of fission energy, development of fast breeder reactors (FBRs) is a necessity. Binary fueled LMFBRs are proposed as the best type for future Brazilian nuclear systems. The inherent safety characteristics are superior to current FBRs and an efficient utilization of the abundant thorium is possible. A first step and a basic tool for the development of FBR technologies is the construction and operation of an experimental fast reactor (EFR). A series of core designs for a 90 MW EFR is studied and possible options and the magnitudes of principal parameters are identified. Flexible modifications of the core and sufficiently high fast fluxes for fuel and materials irradiations appear possible. (Author) [pt

  20. The Maple reactor project

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  1. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    1983-01-01

    The OECD Halden Reactor Project is both the oldest and the only one still in operation of the three major joint undertakings established at the inception of the OECD Nuclear Energy Agency. This publication has been printed in connection with its twenty-fifth anniversary as an international project. After presentation of the history and organization of the project, a thorough description of the past and present activities in the field of fuel performance and process control and surveillance is given. The projects's fuel testing programme is now focuessed on an investigation to define safety margins under normal operations as well as under various kinds of accident situations. Fuel research is also concerned with the characterisation of long term effects with regard to efficiency, operational safety and mapping of reliability and durability in the case of accidents with loss of coolant. In the field of process control and surveillance, research work is directly linked to the use of computers and colour graphics as tools in the control room. A fullscale simulator-based model and experimental control room has been constructed. The first experiments to be carried out in this laboratory will investigate the advantage of analysing alarms before they are presented to the operator. (RF)

  2. Halden reactor project

    International Nuclear Information System (INIS)

    1981-01-01

    Accomplishments realized during 1981 are summarized in this report. Reactor safety considerations continue to be the prime motivation for the Halden fuel programme. A major part of the experimental efforts deal with effects of loss-of-coolant accidents (LOCA) on fuel thermal and dimensional response. Fuel defect mechanisms and probabilities, both during safety related accident sequences and in response to operational transients, are also extensively studied. The programme encompasses current fuel designs as well as design modifications expected to improve performance. The Halden Project is expanding its work on process computer applications with emphasis on operator-process communication and operator guidance systems. This includes human factors experiments, control room layout and design, alarm handling systems and core surveillance. The principal tool being developed for performing realistic experiments in this field is the new control room with a full scope training simulator model of a nuclear power plant

  3. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Blevins, J.D.; Stasko, R.R.

    1989-09-01

    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

  4. Orphee reactor experimental equipment

    International Nuclear Information System (INIS)

    1987-01-01

    Experimental equipment around the ORPHEE reactor is presented. The neutron source; and the spectrometers and sample environment (inelastic and quasi-elastic scattering, elastic scattering, spread scattering, small angle scattering) are described. An experiment proposal and reports guide is supplied [fr

  5. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

    This is the nineteenth annual Report on the OECD Halden Reactor Project, describing activities at the Project during 1978, the last year of the 1976-1978 Halden Agreement. Work continued in two main fields: test fuel irradiation and fuel research, and computer-based process supervision and control. Project research on water reactor fuel focusses on various aspects of fuel behavior under normal, and off-normal transient conditions. In 1978, participating organisations continued to submit test fuel for irradiation in the Halden boiling heavy-water reactor, in instrumented test assemblies designed and manufactured by the Project. Work included analysis of the impact of fuel design and reactor operating conditions on fuel cladding behavior. Fuel performance modelling included characterization of thermal and mechanical behavior at high burn-up, of fuel failure modes, and improvement of data qualification procedures to reduce and quantify error bands on in-reactor measurements. Instrument development yielded new or improved designs for measuring rod temperature, internal pressure, axial neutron flux shape determination, and for detecting cladding defects. Work on computer-based methods of reactor supervision and control included continued development of a system for predictive core surveillance, and of special mathematical methods for core power distribution control

  6. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1978-01-01

    This report summarizes the activities of the OECD Halden Reactor Project for the year 1976. The main items reported on are: a) the process supervision and control which have focused on core monitoring and control, and operator-process communication; b) the fuel performance and safety behavior which have provided data and analytical descriptions of the thermal, mechanical and chemical behavior of fuel under various operating conditions; c) the reactor operations and d) the administration and finance

  7. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    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

  8. Design project of the dosimetry control system in the independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. V

    International Nuclear Information System (INIS)

    1964-01-01

    Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels includes the following: calculations of CO 2 gas activity, design of the dosimetry control system, review of the changes that should be done in the RA reactor building for installing the independent CO 2 loop, specification of the materials with cost estimation, engineering drawings of the system [sr

  9. 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; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 03. Kran za opsluzivanje vertikalnih eksperimentalnih kanala reaktora RA - izrada projekta

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    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.

  10. OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    1988-01-01

    The OECD Halden Reactor project is an agreement between OECD member countries. It was first signed in 1958 and since then regularly renewed every third year. The activities at the Project is centred around the Halden heavy water rector, the HBWR. The reseach programme comprizes studies of fuel performance under various operating conditions, and the application of computers for process control. The HBWR is equipped for exposing fuel rods to temperatures and pressures, and at heat ratings met in modern BWR's and PWR's. A range of in-core instruments are available, permitting detailed measurements of the reactions of the fuel, including mechanical deformations, thermal behaviour, fission gas release, and corrosion. In the area of computer application, the studies of the communication between operator and process, and the surveillance and control of the reactor core, are of particular interst for reactor operation. 1988 represents the 30th year since the Project was started, and this publication is produced to mark this event. It gives and account of the activities and achievements of the Project through the years 1958-1988

  11. ISTC: experimental and technology programs toward novel reactor concepts (review of the ISTC projects and programs)

    Energy Technology Data Exchange (ETDEWEB)

    Tocheny, L.V. [ISTC - International Science and Technology Center, Moscow, Russia, Moscow (Russian Federation)

    2007-07-01

    The ISTC (International Science and Technology Center) is a unique international organization created in Moscow more than twelve years ago by Russia, Usa, E.U. and Japan. Later Korea and Canada, and several CIS countries as well acceded to ISTC. The basic idea behind establishing the ISTC was to support non-proliferation of the mass destruction weapons technologies by re-directing former Soviet weapons scientists to peaceful research thus preventing the drain of dangerous knowledge and expertise from Russia and other CIS countries. Numerous science and technology projects were realized with the ISTC support in different areas, from bio-technologies and environmental problems to all aspects of nuclear studies, including those focused on the development of effective innovative concepts and technologies in the nuclear field, in general, and for improvement of nuclear safety, in particular. The presentation addresses some technical results of the ISTC projects as well as methods and approaches employed by the ISTC to foster close international collaboration and manage projects towards fruitful results. Goals of this presentation are to introduce some of the ISTC programs to international nuclear community to give examples of international cooperation, created in the frames of ISTC, to illustrate the statement of importance of international nuclear experiment as a tool for evidence of new nuclear concepts acceptance, and to make a call for further joint collaboration. (author)

  12. ISTC: experimental and technology programs toward novel reactor concepts (review of the ISTC projects and programs)

    International Nuclear Information System (INIS)

    Tocheny, L.V.

    2007-01-01

    The ISTC (International Science and Technology Center) is a unique international organization created in Moscow more than twelve years ago by Russia, Usa, E.U. and Japan. Later Korea and Canada, and several CIS countries as well acceded to ISTC. The basic idea behind establishing the ISTC was to support non-proliferation of the mass destruction weapons technologies by re-directing former Soviet weapons scientists to peaceful research thus preventing the drain of dangerous knowledge and expertise from Russia and other CIS countries. Numerous science and technology projects were realized with the ISTC support in different areas, from bio-technologies and environmental problems to all aspects of nuclear studies, including those focused on the development of effective innovative concepts and technologies in the nuclear field, in general, and for improvement of nuclear safety, in particular. The presentation addresses some technical results of the ISTC projects as well as methods and approaches employed by the ISTC to foster close international collaboration and manage projects towards fruitful results. Goals of this presentation are to introduce some of the ISTC programs to international nuclear community to give examples of international cooperation, created in the frames of ISTC, to illustrate the statement of importance of international nuclear experiment as a tool for evidence of new nuclear concepts acceptance, and to make a call for further joint collaboration. (author)

  13. Halden reactor project

    International Nuclear Information System (INIS)

    1980-01-01

    The research programme at the Halden Project is focused on the following three areas: 1. In-core behavior of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2. Prediction, surveillance and control of fuel and core performance for which models of fuel and core behavior are developed. 3. Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

  14. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    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

  15. OECD: Halden reactor project

    International Nuclear Information System (INIS)

    1979-01-01

    The work at the Project has continued in the two main fields: test fuel irradiation and fuel research, and computer based process supervision and control. Organizations participating in the Project continue to have their fuel irradiated in the Halden Reactor in instrumented test assemblies designed and manufactured by the Project. The Project's fuel studies continue to focus on specific subjects such as fuel pellet/cladding interaction and heat transfer, fission product release and fuel behavior under loss of coolant conditions. The work on process control and supervision continues in the highly relevant fields of core control and operator-process communication. A system for predictive core control is being developed while special mathematical methods for core power distribution control are being studied. Operator-process communication studies comprise use of computer simulation on colour display as important ingredients, while the work on developing a system for interactive plant disturbance analysis continues

  16. Experimental work and know-how transfer in the framework of the individual project 'Earthquake investigations on the HDR reactor'

    International Nuclear Information System (INIS)

    Corvin, P.; Metz, K.

    1976-03-01

    In the framework of investigations to determine the earthquake resistance of nuclear power plant, carried out by the Applied Nucleonics Company on the HDR reactor in Grosswelzheim/Main, theoretical fundamentals and experimental test procedures in the sense of know-how transfer were elaborated. This enables the corresponding German institutions to carry out similar investigations on their own. In addition ANC carried out safety control measures during the investigations in order to prevent structural damage to the HDR reactor and neighbouring buildings. (orig.) 891 HP [de

  17. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    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 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 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. A 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 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-convertors

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

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

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

  19. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1977-01-01

    The activities of the OECD Halden Reactor Project for the year 1975 are summarized. The period under review is the last year of the three year joint programme which commenced on 1st January, 1973. The main items reported upon are: process supervision and control, test fuel irradiation and fuel research, reactor operations, and administration and finance. The process supervision and control work has been concentrated in two fields: methods development for core surveillance and control, and systems development for operator-process communication. As for fuel test, investigations of the densification phenomenon have continued through irradiations to a maximum of about 16000MWd/tUO 2 . Axial and radial deformations of fuel rods are studied, with the effect of power transients upon the dimensional stability of fuel rods, and fuel-cladding heat transfer and fuel temperature. Thermal models for steady state and transient heat transfer in fuel rods have been developed and the work on thermomechanical models of claddings shows considerable promise

  20. The experimental nuclear reactor: AQUILON

    International Nuclear Information System (INIS)

    Girard, Y.; Koechlin, J.C.; Moreau, J.M.

    1958-01-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [fr

  1. [International Thermonuclear Experimental Reactor support

    International Nuclear Information System (INIS)

    Dean, S.O.

    1990-01-01

    This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

  2. Generation IV reactors: international projects

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

    2003-01-01

    Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

  3. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  4. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1974-01-01

    A major part of the current research programme is devoted to irradiation experiments with a wide variety of heavily instrumented test fuel assemblies, in order to study the thermal and mechanical behavior of fuel rods through in-core measurements, in particular various forms of deformation of cladding and fuel as related to operational conditions and fuel rod design parameters. From these measurements mathematical models are being developed to explain quantitatively the deformation behavior, as well as the thermal properties of the fuel. During 1974, fifty-six instrumented fuel assemblies were irradiated in these experiments. Another major part of the Halden programme is aimed at the development and demonstration of advanced computer-based methods for plant and reactor core control, for safety and protection, and for overall supervision of nuclear power stations. Both the control methods themselves and the associated measurement and control apparatus are being elaborated, and during the year particular progress was made with the ''OPCOM'' process operator communication system

  5. Australia's replacement research reactor project

    International Nuclear Information System (INIS)

    Harris, K.J.

    1999-01-01

    the technical and contractual requirements of the project. A project manager with extensive experience in major defence projects has been contracted and a team of technical and contracts specialists has now completed documentation which has been issued as a Request for Tender to four pre-qualified reactor vendors. The four vendors were selected as a result of a worldwide invitation in mid 1998 for expressions of interest. The tender process is a 'two envelope' system whereby the tenderers are required to submit in one envelope the technical and contractual details of their bid whilst submitting the pricing details in the other envelope. Tenders close on 3 January 2000. A contract is expected to be agreed by mid 2000 with commissioning completed during 2005

  6. Performance of MOX fuel: An overview of the experimental programme of the OECD Halden Reactor Project and review of selected results

    International Nuclear Information System (INIS)

    Wiesenack, W.; McGrath, M.

    2000-01-01

    The OECD Halden Reactor Project has defined an extensive experimental programme related to MOX fuels which is being executed with the objective to provide a performance data base similar to that available for UO 2 . In addition to utilising fresh MOX fuel and re-instrumented segments from LWR irradiations to high burnup, the concept of inert matrix fuel is being addressed. The irradiation in the Halden reactor is performed in rigs allowing steady state, power ramping and cyclic operation. In-pile data are obtained from instrumentation such as fuel centreline thermocouples, pressure transducers, fuel and cladding elongation detectors, and movable gauges for measuring the diametral deformation. Various phenomena can be assessed in this way, e.g. thermal performance, swelling and densification, PCMI and fission gas release. The paper describes the objectives of various experiments and provides examples of temperature, pressure and cladding elongation measurements performed on MOX fuel. Salient results are related to the threshold for the onset of significant fission gas release and the relaxation behaviour in a power ramp-PCMI situation. (author)

  7. Experimental project - Cloud chamber

    International Nuclear Information System (INIS)

    Nour, Elena; Quinchard, Gregory; Soudon, Paul

    2015-01-01

    This document reports an academic experimental project dealing with the general concepts of radioactivity and their application to the cloud room experiment. The author first recalls the history of the design and development of a cloud room, and some definitions and characteristics of cosmic radiation, and proposes a description of the principle and physics of a cloud room. The second part is a theoretical one, and addresses the involved particles, the origins of electrons, and issues related to the transfer of energy (Bremsstrahlung effect, Bragg peak). The third part reports the experimental work with the assessment of a cloud droplet radius, the identification of a trace for each particle (alphas and electrons), and the study of the magnetic field deviation

  8. Design project of the experimental facility for testing uranium creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pavlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    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.

  9. The experimental and technological developments reactor

    International Nuclear Information System (INIS)

    Carbonnier, J.L.

    2003-01-01

    THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

  10. Introduction of the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Kawahara, Hirotaka; Aoyama, Takafumi

    2006-01-01

    The experimental fast reactor JOYO at O-arai Engineering Center of Japan Nuclear Cycle Development Institute is the first liquid metal cooled fast reactor in Japan. This paper describes the plant outline, experiences on the fast reactor technology and test results accumulated through twenty eight years successful operation of JOYO. (author)

  11. Design project of the experimental device for studying the uranium Creep in the reactor; Predprojekat eksperimentalnog uredjaja za ispitivanje CREEP-a urana u reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Odeljenje za reaktorsku eksperimentalnu tehniku, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    The objective of this task was constructing a device for creep testing of uranium, i.e permanent deformation of the irradiated fuel. Deformation results from the influence of thermal neutron flux, temperature, time of irradiation, mechanical stress. This describes two possible technical solutions experimental device for creep testing in the vertical experimental channel and in the horizontal experimental channel of the RA reactor. In addition to the design details, the report covers calculations of heat generation, antireactivity, activation of the materials of the constructed experimental loop, mechanical calculations as well as description of measurements and regulation of the uranium sample temperature.

  12. Construction schedule management of China Experimental Fast Reactor

    International Nuclear Information System (INIS)

    Wang Yue

    2012-01-01

    China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

  13. 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.

  14. Radiation environment of fusion experimental reactor

    International Nuclear Information System (INIS)

    Mori, Seiji; Seki, Yasushi

    1988-01-01

    Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

  15. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  16. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

  17. ITER: the first experimental fusion reactor

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1995-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (''the parties''). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world.The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500MW of fusion power.The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER.The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies.The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. (orig.)

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

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    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)

  19. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-02-01

    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)

  20. Actions needed for RA reactor exploitation - I-IV, Part II, Design project VI-SA 1, Experimental loop for testing the EL-4 reactor fuel elements in the central vertical experimental channel of the RA reactor in Vinca; Radovi za potrebe eksploatacije reaktora RA - I-IV, II Deo, Predprojekat VI-SA 1, Petlja za ispitivanje gorivnih elemenata reaktora EL-4 u centralnom vertikalnom eksperimentalnom kanalu reaktora RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    The objective of installing the VISA-1 loop was testing the fuel elements of the EL-4 reactor. The fuel elements planned for testing are natural UO{sub 2} with beryllium cladding, cooled by CO{sub 2} under nominal pressure of 60 at and temperature 600 deg C. central vertical experimental channel of the RA reactor was chosen for installing a test loop cooled by CO{sub 2}. This report contains the detailed design project of the testing loop with the control system and safety analysis of the planned experiment.

  1. The experimental determination and evaluation of the spectral indices of the IPEN/MB-01 reactor for the IRPhE project

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Fanaro, Leda C.C.B.; Jerez, Rogério

    2013-01-01

    Highlights: • A new experimental method for the spectral index measurements was developed. • The method eliminates the need of calculated correction factors for the case of 28 ρ. • The experimental results were approved for inclusion in the IRPhE handbook. • MCNP5 with ENDF/B-VII.0 described well the experimental data. - Abstract: New experimental results for the spectral indices of the IPEN/MB-01 reactor are presented in this work. The experimental approach considers a new technique for the 28 ρ case which does not require any sort of calculated correction factors. This aspect gave to the IPEN/MB-01 experiment an excellent quality and free of possible bias due to these calculated correction factors. The uncertainty analysis show, even considering the uncertainties of the geometric and material data of the facility, that the final uncertainties are small enough and well understood for a benchmark problem. The experiment was evaluated and included in the IRPhE handbook. The theory/experiment comparison reveals that there is considerable progress in the 238 U nuclear data for application in thermal reactors. The theory/experiment comparison employs the nuclear data libraries ENDF/B-VII.0, ENDF/B-VI.8, JEF3.1, and JENDL3.3 and reveals an excellent agreement for the spectral index 28 ρ * independently of the nuclear data library considered; aspect never found before in several other comparisons. The long term overprediction of the 238 U neutron epithermal capture appears to be eliminated with the improvements in the recent nuclear data libraries. The experimental performed at the IPEN/MB-01 reactor supports the changes in the 238 U nuclear data incorporated in ENDF/B-VII.0 as well as in the other libraries studied in this work. The theory/experiment comparison of 25 δ * and (C8/F) ept show that these spectral indices are in general slightly overpredicted, thus suggesting that the thermal fission cross section of 235 U might be a little bit underestimated

  2. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Yoder, John; Bessler, Nancy J.

    1988-01-01

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  3. JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G. [CEA, DEN, DER, JHR user Facility Interface Manager' , Cadarache, F-13108 St-Paul-Lez-Durance (France); Gonnier, C. [CEA, DEN, DER, SRJH Jules Horowitz Reactor Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Maugard, B. [CEA, DEN, DER, Reactor Department Studies, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under

  4. Opportunities for reactor scale experimental physics

    International Nuclear Information System (INIS)

    1999-01-01

    A reactor scale tokamak plasma will exhibit three areas of physics phenomenology not accessible by contemporary experimental facilities. These are: (1) instabilities generated by energetic alpha particles; (2) self-heating phenomena; and (3) reactor scale physics, which includes integration of diverse physics phenomena, each with its own scaling properties. In each area, selected examples are presented that demonstrate the importance and uniqueness of physics results from reactor scale facilities for both inductive and steady state reactor options. It is concluded that the physics learned in such investigations will be original physics not attainable with contemporary facilities. In principle, a reactor scale facility could have a good measure of flexibility to optimize the tokamak approach to magnetic fusion energy. (author)

  5. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-01-01

    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)

  6. Light water reactor safety research project

    International Nuclear Information System (INIS)

    Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

    1980-07-01

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  7. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    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

  8. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  9. Experimental plasma research project summaries

    International Nuclear Information System (INIS)

    1982-10-01

    The experimental plasma Research Branch has responsibility for developing a broad range of experimental data and new experimental techniques that are required for operating and interpreting present large-scale confinement experiments, and for designing future deuterium-tritium burining facilities. The Branch pursued these objectives by supporting research in DOE laboratories, other Federal laboratories, other Federal laboratories, universities, and private industry. Initiation and renewal of research projects are primarily through submission of unsolicited proposals by these institutions to DOE. Summaries of these projects are given

  10. Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

    Science.gov (United States)

    Shimomura, Y.

    1994-05-01

    The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

  11. ITER [International Thermonuclear Experimental Reactor] reactor building design study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

  12. Experimental Plasma Research project summaries

    International Nuclear Information System (INIS)

    1980-09-01

    This report contains descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Diagnostics; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report

  13. Experimental plasma research project summaries

    International Nuclear Information System (INIS)

    1978-08-01

    This report contans descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Measurements and Instrumentation; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report

  14. Experimental Plasma Research project summaries

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-09-01

    This report contains descriptions of the activities supported by the Experimental Plasma Research Branch of APP. The individual project summaries were prepared by the principal investigators and include objectives and milestones for each project. The projects are arranged in six research categories: Plasma Properties; Plasma Heating; Plasma Diagnostics; Atomic, Molecular and Nuclear Physics; Advanced Superconducting Materials; and the Fusion Plasma Research Facility (FPRF). Each category is introduced with a statement of objectives and recent progress and followed by descriptions of individual projects. An overall budget summary is provided at the beginning of the report.

  15. EL-3 dismantling of an experimental reactor

    International Nuclear Information System (INIS)

    1989-01-01

    The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

  16. MIT pebble bed reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

    2007-03-15

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

  17. MIT pebble bed reactor project

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2007-01-01

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

  18. Aqueous homogeneous suspension reactor project

    International Nuclear Information System (INIS)

    Kersten, J.A.H.

    1976-11-01

    During the period April 1 through September 30, 1976, the energy production of the KSTR reactor was rather small due to problems with the inventory determinations and a decrease of the critical temperature at power operation for which no explanation could be found. A study program in cooperation with the IAEA will be carried out on the status and prospects of thermal breeders. Work done on the investigation of KSTR samples is mentioned briefly

  19. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    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)

  20. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  1. Reactors Project Delivery: The Value of Experiance

    International Nuclear Information System (INIS)

    Stosic, V. Zoran

    2014-01-01

    State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

  2. Prometheus Project Reactor Module Final Report, For Naval Reactors Information

    International Nuclear Information System (INIS)

    MJ Wollman; MJ Zika

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

  3. The International Thermonuclear Experimental Reactor configuration evolution

    International Nuclear Information System (INIS)

    Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

    1989-01-01

    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

  4. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    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)

  5. Dust removal system for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y.; Seki, Y.; Ueda, S.; Aoki, I.

    1995-01-01

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors

  6. Dust removal system for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Ueda, Y.; Takahashi, K.; Oda, Y. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan); Seki, Y.; Ueda, S.; Aoki, I. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)

    1995-12-31

    Development of a dust removal system using static electricity has been conducted. It is envisioned that the system can collect and transport dust under vacuum. In the system, the dust is charged by dielectric polarization and floated by an electrostatic attraction force that is generated by the DC electric field. The dust is then transported by the electric curtain formed by the three-phase AC electric field. Experimental investigation has been conducted to examine the characteristics of the system. Current research results indicate that the dust removal system using static electricity can be used for fusion experimental reactors.

  7. New Production Reactor project-management plan

    International Nuclear Information System (INIS)

    McCrosson, F.J.; Hibbard, L.; Buckner, M.R.

    1982-01-01

    This document provides a project management plan for the first phase of a project to design and build a new production reactor (NPR) at SRP. The design of the NPR is based upon proven SRP heavy water reactor design, with several enhancements such as full containment, moderator detritiation, improved cooling, and modernized control rooms and instrumentation. The first phase of the NPR project includes environmental and safety analyses, preparation of the technical data summary and basic data, site studies, engineering studies, and conceptual design. The project management plan was developed by a 14-member task force comprised of representatives from the Technical Division, the Manufacturing Division, the Departmental Engineer's Office, and the Engineering Department

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-09-27

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

  9. The experimental nuclear reactor: AQUILON; Le reacteur nucleaire experimental: AQUILON

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Koechlin, J C; Moreau, J M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

  10. Modeling a nuclear reactor for experimental purposes

    International Nuclear Information System (INIS)

    Berta, V.T.

    1980-01-01

    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

  11. Remote maintenance for fusion experimental reactor

    International Nuclear Information System (INIS)

    Koizumi, Koichi; Takeda, Nobukazu

    2000-01-01

    Here was introduced on maintenance of reactor core portion operated by remote control among maintenance of the International Thermonuclear Experimental Reactor (ITER) begun on its design since 1988 under international cooperation of U.S.A., Europe, Russia and Japan. Every appliances constructing the reactor core portion is necessary to carry out all of their inspection and maintenance by using remote controlled apparatus because of their radiation due to neutron generated by DT combustion of plasma. For engineering design activity (EDA) in ITER, not only design and development of the remote control appliances but also design under consideration of remote maintenance for from structural design of maintained objective appliances to access method to appliances, transportation and preservation method of radiated matters, and out-reactor maintenance in a hot cell, is now under progress. Here were also reported on basic concept on maintenance and conservation of ITER, maintenance design of diverter and blanket with high maintenance frequency and present state on development of maintenance appliances. (G.K.)

  12. Experimental plasma research project summaries

    International Nuclear Information System (INIS)

    1992-06-01

    This is the latest in a series of Project Summary books going back to 1976 and is the first after a hiatus of several years. They are published to provide a short description of each project supported by the Experimental Plasma Research Branch of the Division of Applied Plasma Physics in the Office of Fusion Energy. The Experimental Plasma Research Branch seeks to provide a broad range of experimental data, physics understanding, and new experimental techniques that contribute to operation, interpretation, and improvement of high temperature plasma as a source of fusion energy. In pursuit of these objectives, the branch supports research at universities, DOE laboratories, other federal laboratories and industry. About 70 percent of the funds expended are spent at universities and a significant function of this program is the training of students in fusion physics. The branch supports small- and medium-scale experimental studies directly related to specific critical plasma issues of the magnetic fusion program. Plasma physics experiments are conducted on transport of particles and energy within plasma and innovative approaches for operating, controlling, and heating plasma are evaluated for application to the larger confinement devices of the magnetic fusion program. New diagnostic approaches to measuring the properties of high temperature plasmas are developed to the point where they can be applied with confidence on the large-scale confinement experiments. Atomic data necessary for impurity control, interpretation of diagnostic data, development of heating devices, and analysis of cooling by impurity ion radiation are obtained. The project summaries are grouped into these three categories of plasma physics, diagnostic development and atomic physics

  13. Preliminary study of a flux converter for experimental reactor

    International Nuclear Information System (INIS)

    Malouch, M.F.

    1998-01-01

    The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

  14. Tokamak Fusion Test Reactor Project

    International Nuclear Information System (INIS)

    1983-01-01

    Where the project's major contracts for buildings and equipment have experienced significant increases in cost, even though the contracts were for fixed prices, IG concluded that a major reason for the increases was the failure to adequately specify requirements prior to awarding the contracts. IG recommended that the Department should develop guidelines on what constitutes an adequate specification for a fixed-price contract and what controls should be placed over change orders. The Assistant Secretary for Management and Administration indicated in his comments to the draft report that an upcoming revision of the Accounting Practices and Procedures Handbook would establish controls over reallocations of construction funds to operating funds. Recommendations also propose that particular attention be given to ensuring that an agreed-upon plan and budget for completing the project is established, that a performance measurement system be implemented by the Princeton Plasma Physics Laboratory, and that improvements be made in the laboratory's Cost/Schedule Performance Report. Improvements are also needed in the quality assurance and safety programs of the Princeton Plasma Physics Laboratory. A number of recommendations from previous quality-assurance and safety reviews, performed by personnel from the Princeton Plasma Physics Laboratory and the Department of Energy, have not been implemented. Comments to the draft report also address these outstanding issues

  15. Conceptual design of fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1985-01-01

    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)

  16. Overview of the OECD Halden reactor project

    International Nuclear Information System (INIS)

    Vitanza, C.

    2000-01-01

    The OECD Halden Reactor Project is an international network dedicated to enhancing the safety and reliability of nuclear power plants. The project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the project's goals and of its technical infrastructure. The paper also contains a brief overview of results from the ongoing programme and of the main issues contemplated for the next three-year programme period (2000-2002). (author)

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

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    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) [pt

  18. Device for rearranging control rods of experimental reactors

    International Nuclear Information System (INIS)

    Louda, J.

    1975-01-01

    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.)

  19. Status of Kijang Resarch Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jee. Y; Kwon, T. H.; Kim, Jun. Y.; Oh, G. B. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The High-flux Advanced Neutron Application Reactor (HANARO) is a multi-purpose reactor in Korea Atomic Energy Research Institute (KAERI) and is being utilized for neutron scattering experiments, material and fuel tests for nuclear power plants, radio-isotope (RI) productions, silicon doping, neutron activation analysis, and neutron radiography. In medical applications, the majority of RIs produced using HANARO are I-131 and Ir-192. Other RIs such as Mo-99 are coming from imports. The self-sufficiency of RI demand becomes an important issue for the public health service in Korea. In this regard the Kijang Research Reactor (KJRR) project was officially launched on the first of April 2012 in need to provide the self-sufficiency of RI demand including Mo-99, increase the neutron transportation doping (NTD) capacity and develop technologies related to the research reactor. When CP is granted, the first excavation is planned to start at the end of this year. In next year, pouring the first concrete and energizing 154kV will follow. In 2018, it is planned to complete utility building construction and reactor building construction.

  20. Fessenheim simulator for OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    Oudot, G.; Bonnissent, B.

    1998-01-01

    A full scope NPP simulator is presently under manufacture by THOMSON TRAINING and SIMULATION (TTandS) in Cergy (France) for the OECD HALDEN REACTOR PROJECT. The reference plant of this simulator is the Fessenheim CP0 PWR power plant operated by the French utility EDF, for which TTandS has delivered a full scope training simulator in mid 1997. The simulator for HALDEN Reactor Project is based on a software duplication of the Fessenheim simulator delivered to EDF, ported on the most recent computers and O.S. available. This paper outlines the main features of this new simulator generation which reaps benefit of the advanced technologies of the SIPA design simulator introduced inside a full scope simulator. This kind of simulator is in fact the synthesis between training and design simulators and offers therefore added technical capabilities well suited to HALDEN needs. (author)

  1. The PALLAS research and isotope reactor project status

    International Nuclear Information System (INIS)

    Van Der Schaaf, B.; De Jong, P.

    2010-01-01

    In the European Union the first generation research reactors is nearing their end of life condition. Several committees recommend a comprehensive set of reactors in the EU, amongst them the replacement for the HFR research and isotope reactor in Petten: PALLAS. The business case for PALLAS supports a future for a research and isotope reactor in Petten as a perfect fit for the future EU set of test reactors. The tender for PALLAS started in 2007, following the EU rules for tendering complex objects with the competitive dialogue. This procedure involved an extensive consultation phase between individual tendering companies and NRG, resulting in definitive specifications in summer 2008. The evaluation of offers, including conceptual designs, took place in summer 2009. At present NRG is still active in the acquisition of the funding for the project. The licensing path has been started in autumn 2009 with a initiation note on the environmental impact assessment, EIA. The public hearings held in the lead to the advice from the national EIA committee for the approach of the assessment. The PALLAS project team in Petten will guide the design and build processes. It is also responsible for the licensing of the building and operation of PALLAS. The team also manages the design and construction for the infrastructure, such as cooling devices, including remnant heat utilization, and utility provisions. A particular responsibility for the team is the design and construction of experimental and isotope capsules, based on launch customer requirements. (author)

  2. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  3. JANUS reactor d and d project

    International Nuclear Information System (INIS)

    Fellhauer, C. R.

    1998-01-01

    Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

  4. RA research reactor - properties and experimental capabilities

    International Nuclear Information System (INIS)

    Milosevic, M.; Martinc, R.

    1978-01-01

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described [sr

  5. Alternative methodology for irradiation reactor experimental shielding calculation

    International Nuclear Information System (INIS)

    Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

    1996-01-01

    Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

  6. Plant experience of experimental fast reactor 'Joyo'

    International Nuclear Information System (INIS)

    1982-01-01

    The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

  7. The Jules Horowitz Reactor project, a driver for revival of the research reactor community

    International Nuclear Information System (INIS)

    Pere, P.; Cavailler, C.; Pascal, C.

    2010-01-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first-rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10 14 n.cm -2 /sec -1 E>1 MeV in core and 3,6x10 14 n.cm -2 /sec -1 E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960s, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items:accommodation of a broad experimental domain; involvement by international partners making in-kind contributions to the project; ? development of components critical to safety and performance; the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and; tools to carry out the project, including computer codes

  8. The Jules Horowitz reactor project, a driver for revival of the research reactor community

    Energy Technology Data Exchange (ETDEWEB)

    Pere, P.; Cavailler, C.; Pascal, C. [AREVA TA, CEA Cadarache - Etablissement d' AREVA TA - Chantier RJH - MOE - BV2 - BP no. 9 - 13115 Saint Paul lez Durance (France); CS 50497 - 1100, rue JR Gauthier de la Lauziere, 13593 Aix en Provence cedex 3 (France)

    2010-07-01

    The first concrete of the nuclear island for the Jules Horowitz Reactor (JHR) was poured at the end of July 2009 and construction is ongoing. The JHR is the largest new platform for irradiation experiments supporting Generation II and III reactors, Generation IV technologies, and radioisotope production. This facility, composed of a unique grouping of workshops, hot cells and hot laboratories together with a first -rate MTR research reactor, will ensure that the process, from preparations for irradiation experiments through post-irradiation non-destructive examination, is completed expediently, efficiently and, of course, safely. In addition to the performance requirements to be met in terms of neutron fluxes on the samples (5x10{sup 14} n.cm{sup -2}/sec{sup -1} E> 1 MeV in core and 3,6x10{sup 14} n.cm{sup -2}/sec{sup -1} E<0.625 eV in the reflector) and the JHR's considerable irradiation capabilities (more than 20 experiments and one-tenth of irradiation area for simultaneous radioisotope production), the JHR is the first MTR to be built since the end of the 1960's, making this an especially challenging project. The presentation will provide an overview of the reactor, hot cells and laboratories and an outline of the key milestones in the project schedule, including initial criticality in early 2014 and radioisotope production in 2015. This will be followed by a description of the project organization set up by the CEA as owner and future operator and AREVA TA as prime contractor and supplier of critical systems, and a discussion of project challenges, especially those dealing with the following items: - accommodation of a broad experimental domain, - involvement by international partners making in-kind contributions to the project, - development of components critical to safety and performance, - the revival of engineering of research reactors and experimental devices involving France's historical players in the field of research reactors, and

  9. The Brazilian Multipurpose Reactor (RMB) Project

    Energy Technology Data Exchange (ETDEWEB)

    Perrotta, Jose Augusto [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2012-07-01

    Full text: The Plan of Action on Science, Technology and Innovation (PACT 2007-2010) of the Ministry of Science Technology and Innovation (MCTI), aligned to the governmental strategies for the Brazilian Nuclear Program, established as a goal the study and definition of the Brazilian Multipurpose Reactor (RMB). The RMB research reactor is designed to perform three main functions: radioisotope production for medicine, industry, agriculture and environmental applications; fuel and material irradiation testing in support to the Brazilian nuclear energy program; and to provide neutron beams for scientific and applied research. The main project facilities are: nuclear pool type reactor with a flux level compatible to the multipurpose uses; hot cells laboratory for Mo-99 and I-131 processing; hot cells laboratory for radioisotope processing; hot cells laboratory for irradiated material post irradiation analysis; neutron beams laboratory building with scientific equipment and instrumentation for researching; radiochemistry laboratory; radioactive waste treatment facility; support laboratories for operation and researching; and buildings for researchers and operators. This speech presents the RMB project status, giving some technical and management details on its development and its future perspectives for new jobs in research activities for the Brazilian technical and scientific community. (author)

  10. The Brazilian Multipurpose Reactor (RMB) Project

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    2012-01-01

    Full text: The Plan of Action on Science, Technology and Innovation (PACT 2007-2010) of the Ministry of Science Technology and Innovation (MCTI), aligned to the governmental strategies for the Brazilian Nuclear Program, established as a goal the study and definition of the Brazilian Multipurpose Reactor (RMB). The RMB research reactor is designed to perform three main functions: radioisotope production for medicine, industry, agriculture and environmental applications; fuel and material irradiation testing in support to the Brazilian nuclear energy program; and to provide neutron beams for scientific and applied research. The main project facilities are: nuclear pool type reactor with a flux level compatible to the multipurpose uses; hot cells laboratory for Mo-99 and I-131 processing; hot cells laboratory for radioisotope processing; hot cells laboratory for irradiated material post irradiation analysis; neutron beams laboratory building with scientific equipment and instrumentation for researching; radiochemistry laboratory; radioactive waste treatment facility; support laboratories for operation and researching; and buildings for researchers and operators. This speech presents the RMB project status, giving some technical and management details on its development and its future perspectives for new jobs in research activities for the Brazilian technical and scientific community. (author)

  11. Project based learning for reactor engineering education

    International Nuclear Information System (INIS)

    Narabayashi, Tadashi; Tsuji, Masashi; Shimazu, Yoichiro

    2009-01-01

    Trial in education of nuclear engineering in Hokkaido University has proved to be quite attractive for students. It is an education system called Project Based Learning (PBL), which is not based on education by lecture only but based mostly on practice of students in the classroom. The system was adopted four years ago. In the actual class, we separated the student into several groups of the size about 6 students. In the beginning of each class room time, a brief explanations of the related theory or technical bases. Then the students discuss in their own group how to precede their design calculations and do the required calculation and evaluation. The target reactor type of each group was selected by the group members for themselves at the beginning of the semester as the first step of the project. The reactor types range from a small in house type to that for a nuclear ship. At the end of the semester, each group presents the final design. The presentation experience gives students a kind of fresh sensation. Nowadays the evaluation results of the subject by the students rank in the highest in the faculty of engineering. Based on the considerations above, we designed the framework of our PBL for reactor engineering. In this paper, we will present some lessons learned in this PBL education system from the educational points of view. The PBL education program is supported by IAE/METI in Japan for Nuclear Engineering Education. (author)

  12. Conceptual design of Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Tone, T.; Fujisawa, N.

    1983-01-01

    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/m 2 , 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

  13. TRIGA reactor as an experimental tool

    Energy Technology Data Exchange (ETDEWEB)

    Nahrul Khair bin Alang Mohammad Rashid (PUSPATI, Selangor (Malaysia))

    1981-01-01

    Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor.

  14. Triga reactor as an experimental tool

    International Nuclear Information System (INIS)

    Nahrul Khair bin Alang Mohammad Rashid

    1981-01-01

    Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor

  15. Advanced Reactor Technology/Energy Conversion Project FY17 Accomplishments.

    Energy Technology Data Exchange (ETDEWEB)

    Rochau, Gary E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    The purpose of the ART Energy Conversion (EC) Project is to provide solutions to convert the heat from an advanced reactor to useful products that support commercial application of the reactor designs.

  16. Experimental techniques applied at the RB reactor

    International Nuclear Information System (INIS)

    Markovic, H.; Takac, S.; Sotic, O.; Dimitrijevic, Z.

    1979-12-01

    This paper contains a brief description of research and operations at the RB reactor which are concerned with experiments and results of measuring typical reactor parameters, neutron characteristics as well as parameters related to reactor operation and utilization. Annex contains a list of relevant original papers and publications [sr

  17. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

    International Nuclear Information System (INIS)

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-01-01

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, 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 (SPF) 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 SPF

  18. 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)

    Tassan, S.

    1978-01-01

    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

  19. OECD high temperature reactor project Dragon

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented concerning the Dragon reactor support studies and fuel irradiation programs, HTGR and fuel graphite studies, primary circuit materials, reactor safety evaluation, and administration

  20. Media and Australia's replacement reactor project

    International Nuclear Information System (INIS)

    Keenan, Pamela

    2001-01-01

    In September 1997, the Commonwealth Government of Australia announced a proposal to build a replacement nuclear research reactor at Lucas Heights in Sydney. Extensive public consultation, parliamentary debate and independent reports were prepared to ensure that the new facility would meet strict international requirements, national safety and environmental standards, and performance specifications servicing the needs of Australia - for decades to come. On 6 June 2000, Argentine company INVAP SE was announced as the preferred tenderer. In July 2000 contracts were signed between INVAP and the Australian Nuclear Science and Technology Organisation for the construction the replacement reactor, due to be completed in 2005. In order to retain a strong local presence, INVAP undertook a joint venture with two of Australia's foremost heavy construction businesses. Briefly the new research reactor will be a replacement for the ageing Australian Reactor (HIFAR). Nuclear science and technology, in Australia, is no stranger to media controversy and misinformation. Understandably the announcement of a preferred tenderer followed by the signing of contracts, attracted significant national and international media attention. However in the minds of the media, the issue is far from resolved and is now a constant 'news story' in the Australian media. Baseless media stories have made claims that the project will cost double the original estimates; question the credibility of the contractors; and raise issues of international security. The project is currently linked with Australia's requirements for long term nuclear waste management and there has been an attempt to bring national Indigenous People's issues into play. Some of these issues have been profiled in the press internationally. So, just to set the record straight and give you an appropriate impression of what's 'really happening' I would like to highlight a few issues, how ANSTO dealt with these, and what was finally reported

  1. Jules Horowitz reactor - presentation of the project and progress report

    International Nuclear Information System (INIS)

    Ballagny, A.; Frachet, S.; Rommens, M.; Guigon, B.; Minguet, J.L.; Dupuy, J.P.; Leydier, C.

    2001-01-01

    The RJH project was launched by CEA some years ago, with the objective to replace, after 2010, the material testing reactors of the previous generation. The objectives are also: to realise a significant step in term of performances, to ensure a high flexibility of the design, in order to host in the future new experiments, which are not completely defined at the project stage, to reach a high level of safety, according to the best current practice. After a summary of the main experimental objectives of the facility, the present paper deals with a detailed technical presentation of the project, resulting from preliminary design studies. The following topics are covered successively: the main functionalities, the resulting design options and technical solutions, the layout of the nuclear facility. (orig.)

  2. First preliminary design of an experimental fusion reactor

    International Nuclear Information System (INIS)

    1977-09-01

    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 Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

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

    International Nuclear Information System (INIS)

    Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

    1979-06-01

    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)

  4. Sandia Pulse Reactor-IV Project

    International Nuclear Information System (INIS)

    Reuscher, J.A.

    1983-01-01

    Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

  5. Advanced nuclear reactor public opinion project

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  6. Advanced nuclear reactor public opinion project

    International Nuclear Information System (INIS)

    Benson, B.

    1991-01-01

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions

  7. Calculation of neutron spectra in the reactor cell of the RA experimental reactor in Vinca

    International Nuclear Information System (INIS)

    Bosevski, T.; Altiparmakov, D.; Marinkovic, N.

    1974-01-01

    In the frame of neutron properties of RA experimental reactor the study of energy neutron spectra in the reactor cell are planned. Complex reactor cell geometry, nine cylindrical regions causes high space-energy variations of neutron flux with a significant gradient both in energy and space variables. Treatment of such a complex problem needs adequate methodology which ensures reliable results and control of accuracy. This paper describes in detail the method for calculating group constants based on lattice cell calculation for the need of calculation of reactor core parameters. In 26 group approximation for the energy region from 0 - 10.5 MeV, values of neutron spectra are obtained in 18 space points chosen to describe, with high accuracy, integral reactor cell parameters of primary importance for the reactor core calculation. Obtained space-energy distribution of neutron flux in the reactor cell is up to now unique in the study of neutron properties of Ra reactor [sr

  8. The aqueous homogeneous suspension reactor project

    International Nuclear Information System (INIS)

    1975-01-01

    The power of the KSTR reactor has been increased up to 200 kW in the fourth quarter of 1974. A description is given of the behaviour of the reactor at increased power level, safety aspects concerned with this new level, the operation of the reactor, instrumental behavior and mechanical behavior. Irradiation investigation of two types of fuels are reported and results are presented. Progress made on the conceptual design of a 250 MWe suspension reactor is described

  9. Experimental Equipment for Physics Studies in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernander, G; Blomberg, P E; Dubois, P O

    1967-03-15

    Comprehensive physics measurements were carried out in connection with the start up of the Agesta reactor. For this purpose special experimental equipment was constructed and installed in the reactor. Parts of this were indispensable and/or time-saving for the reactivity control during the core build-up period and during the first criticality studies. This report gives mainly a detailed description of the experimental equipment used, but also the relevant physics background and the experience gained during the performance.

  10. Brazilian Irradiation Project: CAFE-MOD1 validation experimental program

    International Nuclear Information System (INIS)

    Mattos, Joao Roberto Loureiro de; Costa, Antonio Carlos L. da; Esteves, Fernando Avelar; Dias, Marcio Soares

    1999-01-01

    The Brazilian Irradiation Project whose purpose is to provide Brazil with a minimal structure to qualify the design, fabrication and quality procedures of nuclear fuels, consists of three main facilities: IEA-R1 reactor of IPEN-CNEN/SP, CAFE-MOD1 irradiation device and a unit of hot cells. The CAFE-MOD1 is based on concepts successfully used for more than 20 years in the main nuclear institutes around the world. Despite these concepts are already proved it should be adapted to each reactor condition. For this purpose, there is an ongoing experimental program aiming at the certification of the criteria and operational limits of the CAFE-MOD1 in order to get the allowance for its installation at the IEA-R1 reactor. (author)

  11. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  12. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    Bess, John D.; Gullifor, Jim

    2015-03-01

    The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

  13. Emergency reactor cooling systems for the experimental VHTR

    International Nuclear Information System (INIS)

    Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

    1983-03-01

    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)

  14. Measurements at the RA Reactor related to the VISA-2 project - Part 1, Start-up of the RA reactor and measurement of new RA reactor core parameters

    International Nuclear Information System (INIS)

    Markovic, H.

    1962-07-01

    The objective of the measurements was determining the neutron flux in the RA reactor core. Since the number of fuel channels is increased from 56 to 68 within the VISA-2 project, it was necessary to attain criticality of the RA reactor and measure the neutron flux properties. The 'program of RA reactor start-up' has been prepared separately and it is included in this report. Measurements were divided in two phases. First phase was measuring of the neutron flux after the criticality was achieved but at zero power. During phase two measurements were repeated at several power levels, at equilibrium xenon poisoning. This report includes experimental data of flux distributions and absolute values of the thermal and fast neutron flux in the RA reactor experimental channels and values of cadmium ratio for determining the neutron epithermal flux. Data related to calibration of regulatory rods for cold un poisoned core are included [sr

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

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L C

    1991-11-15

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

  16. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  17. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

  18. Status of the RA research reactor decommissioning project

    International Nuclear Information System (INIS)

    Ljubenov, V.; Nikolic, D.; Pesic, M.; Milosevic, M.; Kostic, Lj.; Steljic, M.; Sotic, O.; Antic, D. . E-mail address of corresponding author: vladan@vin.bg.ac.yu; Ljubenov, V.)

    2005-01-01

    The 6.5 MW heavy water RA research reactor at the VINCA Institute of Nuclear Sciences operated from 1959 to 1984. After 18 years of extended shutdown in 2002 it was decided that the reactor shutdown should be final. Preliminary decommissioning activities have been initiated by the end of 2002 under the Technical Co-operation Programme of the International Atomic Energy Agency. The objective of the project is to implement safe, timely and cost-effective decommissioning of the RA reactor up to unrestricted use of the site. Decommissioning project is closely related to two other projects: Safe Removal of the RA Reactor Spent Nuclear Fuel and Radioactive Waste Management in VINCA Institute. The main phases of the project include preparation of the detailed decommissioning plan, radiological characterization of the reactor site, dismantling and removal of the reactor components and structures, decontamination, final radiological site survey and the documentation of all the activities in order to obtain the approval for unrestricted use of the facility site. In this paper a review of the activities related to the preparation and realization of the RA reactor decommissioning project is given. Status of the project's organizational and technical aspects as for July 2004 are presented and plans for the forthcoming phases of the project realization are outlined. (author)

  19. The aqueous homogeneous suspension reactor project

    International Nuclear Information System (INIS)

    1976-01-01

    During 1975, reactor power has been increased to the design power of 1000 KW, whereas power fluctuations show a decrease with increased mean power. Operation experience with the reactor and associated instrumentation during 1975 is described. The results of the experiments done for fuel irradiations and investigations in the KSTR fuel, mainly to determine the amount of erosion products on the fuel, are described. Concerning the safety of operation of the KSTR reactor, several actions had to be taken mainly to replace or repair components or instrumentation of the reactor. Radiological safety and radioactivity discharges during 1975 are reported

  20. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  1. Independent CO2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Task 2.50.05

    International Nuclear Information System (INIS)

    Stojic, M.; Pavicevic, M.

    1964-01-01

    This report contains the following volumes V and VI of the Project 'Independent CO 2 loop for cooling the samples irradiated in RA reactor vertical experimental channels': Design project of the dosimetry control system in the independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels, and Safety report for the Independent CO 2 loop for cooling the samples irradiated in the RA reactor vertical experimental channels [sr

  2. IAEA fast reactor knowledge preservation initiative. Project focus: KNK-II reactor, Karlsruhe, Germany

    International Nuclear Information System (INIS)

    2004-08-01

    This Working Material (including the attached CD-ROM) documents progress made in the IAEA's initiative to preserve knowledge in the fast reactor domain. The brochure describes briefly the context of the initiative and gives an introduction to the contents of the CD-ROM. In 2003/2004 a first focus of activity was concentrated on the preservation of knowledge related to the KNK-II experimental fast reactor in Karlsruhe, Germany. The urgency of this project was given by the impending physical destruction of the installation, including the office buildings. Important KNK-II documentation was brought to safety and preserved just in time. The CD-ROM contains the full texts of 264 technical and scientific documents describing research, development and operating experience gained with the KNK-II installation over a period of time from 1965 to 2002, extending through initial investigations, 17 years of rich operating experience, and final shutdown and decommissioning. The index to the documents on the CD-ROM is printed at the end of this booklet in chronological order and is accessible on the CD by subject index and chronological index. The CD-ROM contains in its root directory also the document 'fr c lassification.pdf' which describes the classification system used for the present collection of documents on the fast reactor KNK-II

  3. Pre-design stage of the intermediate heat exchanger for experimental fast reactor

    International Nuclear Information System (INIS)

    Luz, M.; Borges, E.M.; Braz Filho, F.A.; Hirdes, V.R.

    1986-09-01

    This report presents the outlines of a thermal-hydraulic calculation procedure for the pre-design stage of the Intermediate Heat Exchanger for a 5 MW Experimental Fast Reactor (EFR), which can be used in other similar projects, at the same stage of evolution. Heat transfer and heat loss computations for the preliminary design of the heat exchanger are presented. (author) [pt

  4. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  5. A sustainability analysis of the Brazilian multipurpose reactor project

    International Nuclear Information System (INIS)

    Obadia, I.J.; Perrotta, J.A.

    2010-01-01

    The project of a new research reactor in Brazil for radioisotope production, support of the nuclear energy program and scientific research has received a positive sign of the government and is starting to be developed by the Brazilian Commission of Nuclear Energy. International Atomic Energy Agency points out that the implementation of a new research reactor is a major undertaking for a country, requiring an analysis to identify to which extent the conditions of the national nuclear program are proper and adequate to lead to a sustainable research reactor life cycle. This paper introduces the Brazilian Multipurpose Reactor Project (RMB) and describes the sustainability analysis performed, which has shown that the national nuclear infrastructure presents a very favourable condition to the implementation of the RMB project as well as to provide a sustainable life cycle for this new research reactor. (author)

  6. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  7. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  8. Reactor-specific spent fuel discharge projections, 1987-2020

    International Nuclear Information System (INIS)

    Walling, R.C.; Heeb, C.M.; Purcell, W.L.

    1988-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs

  9. Reactor-specific spent fuel discharge projections: 1986 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Walling, R.C.; Purcell, W.L.

    1987-03-01

    The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No new orders with extended burnup, (2) No new orders with constant burnup, (3) Upper reference (which assumes extended burnup), (4) Upper reference with constant burnup, and (5) Lower reference (which assumes extended burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel. 6 refs., 8 figs., 8 tabs

  10. Aspects of 238Pu production in the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

    2005-01-01

    Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

  11. RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

  12. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  13. Experimental facility of innovative types as the laboratory analog of research reactor experimental device

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Zabud'ko, A.N.; Kremenetskij, A.K.; Nikolaev, A.N.; Trykov, L.A.

    1991-01-01

    The paper analyses capability of creating laboratory analogs of complex experimental facilities at research reactors utilizing power radionuclide neutron sources fabricated in industrial conditions. Some experimental and calculational investigations of neutron-physical characteristics are presented, which have been attained at the RIZ research reactor laboratory analog. Experimental results are supplemented by calculational investigations, fulfilled by means of the BRAND three-dimensional computational complex and the ROZ-6 one-dimensional program. 4 refs.; 3 figs

  14. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

    Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

    1985-11-01

    The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

  15. Experimental utilization of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Bitelli, U. d'Utra; Santos, A. dos; Jerez, R.; Diniz, R.; Fanaro, L.C.C.B.; Abe, A.Y.; Moreira, J.M.L.; Fer, N.; Giada, M.R.; Fuga, R.

    2003-01-01

    This paper aims to show the experimental utilization of the IPEN/MB-01 nuclear reactor during the last fourteen years. The IPEN/MB-01 is a zero-power critical assembly specially designed to measure integral and differential reactor physics parameters to validate calculational methodologies and related nuclear data libraries. Experiments involving determination of spectral indices, critical mass, relative abundance of delayed neutrons, the inversion point of the isothermal reactivity coefficient and burnable poison are considered the most important experiments. Current experiments at IPEN/MB-01 reactor are also commented. (author)

  16. Instrumentation and control improvements at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

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

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I ampersand 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 ampersand 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 ampersand C systems of the next generation of liquid metal reactor (LMR) plants

  17. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

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

  18. Research reactor RB, technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1978-01-01

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

  19. The brief introduction to decommissioning of nuclear reactor projects

    International Nuclear Information System (INIS)

    Zhao Shixin

    1991-01-01

    The basic concept and procedure of the decommissioning of nuclear reactor project and the three stages of decommissioning defined by IAEA are introduced. The main work of decommissioning of nuclear reactor are as following: (1) the documentary and technological preparation; (2) the site preparation of decommissioning project; (3) the dismantling of equipment piping system and components; (4) the decontamination of the piping system before and after decomminssioning; (5) the storage and disposal of the operational and decommissioning waste

  20. The brief introduction to decommissioning of nuclear reactor projects

    Energy Technology Data Exchange (ETDEWEB)

    Shixin, Zhao [Beijing Inst. of Nuclear Engineering (China)

    1991-08-01

    The basic concept and procedure of the decommissioning of nuclear reactor project and the three stages of decommissioning defined by IAEA are introduced. The main work of decommissioning of nuclear reactor are as following: (1) the documentary and technological preparation; (2) the site preparation of decommissioning project; (3) the dismantling of equipment piping system and components; (4) the decontamination of the piping system before and after decomminssioning; (5) the storage and disposal of the operational and decommissioning waste.

  1. Development Project of Supercritical-water Cooled Power Reactor

    International Nuclear Information System (INIS)

    Kataoka, K.; Shiga, S.; Moriya, K.; Oka, Y.; Yoshida, S.; Takahashi, H.

    2002-01-01

    A Supercritical-water Cooled Power Reactor (SCPR) development project (Feb. 2001- Mar. 2005) is being performed by a joint team consisting of Japanese universities and nuclear venders with a national fund. The main objective of this project is to provide technical information essential to demonstration of SCPR technologies through concentrating three sub-themes: 'plant conceptual design', 'thermohydraulics', and 'material and water chemistry'. The target of the 'plant conceptual design sub-theme' is simplify the whole plant systems compared with the conventional LWRs while achieving high thermal efficiency of more than 40 % without sacrificing the level of safety. Under the 'thermohydraulics sub-theme', heat transfer characteristics of supercritical-water as a coolant of the SCPR are examined experimentally and analytically focusing on 'heat transfer deterioration'. The experiments are being performed using fron-22 for water at a fossil boiler test facility. The experimental results are being incorporated in LWR analytical tools together with an extended steam/R22 table. Under the 'material and water chemistry sub-theme', material candidates for fuel claddings and internals of the SCPR are being screened mainly through mechanical tests, corrosion tests, and simulated irradiation tests under the SCPR condition considering water chemistry. In particular, stress corrosion cracking sensitivity is being investigated as well as uniform corrosion and swelling characteristics. Influences of water chemistry on the corrosion product characteristics are also being examined to find preferable water condition as well as to develop rational water chemistry controlling methods. (authors)

  2. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  3. Reactor-specific spent fuel discharge projections: 1985 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

    1986-09-01

    The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel

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

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    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.)

  5. The Management of Projects and Product Experimentation

    DEFF Research Database (Denmark)

    Lorenzen, Mark; Frederiksen, Lars

    2005-01-01

    The paper analyses management of product innovation in project-based industries, offering a view on management not only of firms, but also of markets. It first argues that projects are prominent in industries where the nature of consumer demand means that product innovation takes place...... as experimentation. Then, the paper argues that if skills needed for projects are very diverse and projects are complex, there are few internal managerial economies of projects, and the scope for management then transcends the boundaries of firms. In these cases, markets become organized in combinations of people......, contracts, and other institutions, in order to facilitate the coordination of market-based projects. While contracts play a role, a continuous, active role of knowledgeable managers (leaders and boundary spanners) is also often necessary. Such managers --- and thus (core parts of) whole industries...

  6. Experimental determination of neutron temperature distribution in reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.

    1965-12-01

    This paper describes theoretical preparation of the experiment for measuring neutron temperature distribution at the RB reactor by activation foils. Due to rather low neutron flux Cu and Lu foil were irradiated for 4 days. Special natural uranium fuel element was prepared to enable easy removal of foils after irradiation. Experimental device was placed in the reactor core at half height in order to measure directly the mean neutron density. Experimental data of neutron temperature distribution for square lattice pitch 16 cm are presented with mean values of neutron temperature in the moderator, in the fuel and on the fuel element surface

  7. The SCARABEE experimental fast reactor safety programme already completed

    International Nuclear Information System (INIS)

    Schmitt, A.P.; Teague, H.; Heusener, G.

    1979-08-01

    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

  8. Experimental measurement of zero power reactor transfer function

    International Nuclear Information System (INIS)

    Liang Shuhong

    2011-01-01

    In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

  9. Training experience at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report

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

    International Nuclear Information System (INIS)

    Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

    1987-09-01

    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)

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

    International Nuclear Information System (INIS)

    Jammes, Ch.

    2010-07-01

    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)

  12. Technical modifications and management innovations in exporting nuclear reactor projects

    International Nuclear Information System (INIS)

    Mao Xiaoming; Qin Xijiu; Ding Hu; Xue Zhaoqun; Wen Shengjun

    2009-01-01

    As a main channel for the foreign economic cooperation of China nuclear industry, China Zhongyuan Engineering Corporation (CZEC) has been constantly engaged in technical modifications and management innovations in its exporting nuclear reactor projects. In the implementation of heavy water research reactor contract in Algeria, CZEC had established a complete and adequate design standards system in compliance with the international standards, and made significant modifications to the reference reactor in the aspects of reactor power and reactor safety, solved quite some technical issues which-affected the reactor technical performance. The modifications and improvements enabled the technical parameters, safety features, reactor multipurpose application to attain to the advanced level in the world. In the 300 MWe PWR NPPs in Pakistan, safety features had been updated in line with upgrading regulatory requisites. The design philosophy and technology application demonstrated CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC in promoting China made equipment items and components exportation. (authors)

  13. Selected power reactor projects in Canada and the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-11-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

  14. Selected power reactor projects in Canada and the United States of America

    International Nuclear Information System (INIS)

    1964-01-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

  15. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  16. Experimental research of reactor core flooding

    International Nuclear Information System (INIS)

    Blaha, V.; Kotrnoch, J.; Krett, V.

    1978-01-01

    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)

  17. Thorium-based Molten Salt Reactor (TMSR) project in China

    International Nuclear Information System (INIS)

    Dai, Zhimin; Liu, Wei

    2013-01-01

    Making great efforts in development of nuclear energy is one of the long-term-plan in China's energy strategies. The advantages of Thorium-based nuclear energy are: rich resource in nature, less nuclear waste, low toxicity, nuclear non-proliferation and so on. Furthermore, China is a country with abundant thorium, thus it is necessary to develop the Thorium-based Molten Salt Reactor (TMSR) in China. Shanghai Institute of Applied Physics, Chinese Academy of Sciences (SINAP) had designed and constructed the first China's light-water reactor and developed a zero-power thorium-based molten salt reactor successfully in the early 1970s. The applied research project 'thorium molten salt reactor nuclear power system' by SINAP together with several other institutes had been accepted and granted by China government in 2011. The whole project has been divided into three stages: Firstly, built a 2 MW-zero-power high temperature solid molten salt reactor in 2015 and a 2 MW-zero-power high temperature liquid molten salt reactor in 2017. Secondly, in 2020 built a 10 MW high temperature liquid molten salt reactor. Thirdly, on the base of previous work, a 100 MW high temperature molten salt reactor should be achieving in 2030. After more than one years of efforts, a high quality scientific research team has been formed, which is able to design the molten salt reactor, the molten salt loop and related key equipment, the systems of molten salt preparation, purification and the radioactive gas removal. In the past one year, the initial physical design of high temperature molten salt reactor has been completed; the nuclear chemistry and radiation chemical laboratory has been built, a high temperature salt (HTS) loop and radioactive gas removal experiment device system have been successfully developed and constructed. Further, the preliminary study on reactor used carbon-carbon composite material has been investigated. (author)

  18. 105-B Reactor museum feasibility assessment (Phase 2) project

    International Nuclear Information System (INIS)

    Heckel, R. P.

    2000-01-01

    This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

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

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

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

  20. Regulatory Framework for Controlling the Research Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2009-01-01

    Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

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

    Energy Technology Data Exchange (ETDEWEB)

    GERBER MS

    2009-04-28

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

  2. An overview of the fuels and materials testing programme at the OECD Halden Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-08-01

    The fuels and materials testing programme of the OECD Halden Reactor Project is aimed at investigations of fuel and cladding properties at high burnup, water chemistry effects and in-core materials ageing problems. For the execution of this programme, different types of irradiation rigs and experimental facilities providing typical power reactors conditions are available. Data are obtained from in-core sensors developed at the Halden Project; these are shortly described. An overview of the current test programme and the scope of the following years are briefly presented. (author). 5 refs, 3 figs.

  3. Revised design for the Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1977-03-01

    A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

  4. Neutronic scoping studies for the tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

    1976-02-01

    One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

  5. Projecting regulatory expectations for advanced reactor designs

    International Nuclear Information System (INIS)

    Viktorov, A.

    2011-01-01

    This paper explores the overarching safety principles that will likely guide the safety design of advanced reactor technologies. As will be shown, the already established safety framework provides a solid foundation for the safety design of future nuclear power plants. As a specific example, the principle of 'proven technology' is presented in greater detail and its implications for a novel technology are discussed. Research, modeling and prototyping are shown to be components in satisfying this principle. While the fundamental safety principles are in place, their interpretation may depend both on the considered technology as well as the national context. Thus, the regulatory authority will need to be engaged, at an appropriate stage of the technology development, in specifying the regulatory requirements that will have to be met for a specific reactor design. (author)

  6. Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1988-01-01

    The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

  7. Tasks related to increase of RA reactor exploitation and experimental potential, 04. Device for transport of radioactive reactor channels and semi channels of the RA reactor, design project (I-III) Part II, Vol. II; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 04. Uredjaj za transport aktivnih tehnoloskih kanala I semikanala reaktora RA - izrada projekta (I-III), II Deo, Album II

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    This second volume includes calculations of the main components of the transporter, description of the mechanical part of the transporter and the engineering drawing of the device for transport of radioactive reactor channels and semi channels of the RA reactor.

  8. IRPhE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents

    International Nuclear Information System (INIS)

    2004-01-01

    Description: The DRAGON Reactor Experiment (DRE): The first demonstration High temperature gas reactor (HTGR) was built in the 1960's. Thirteen OECD countries began a project in 1959 to build an experimental reactor known as Dragon at Winfrith in the UK. The reactor - which operated successfully between 1966 and 1975 - had a thermal output of 20 MW and achieved a gas outlet temperature of 750 deg. C. The High Temperature Reactor concept, if it justified its expectations, was seen as having its place as an advanced thermal reactor between the current thermal reactor types such as the PWR, BWR, and AGR and the sodium cooled fast breeder reactor. It was expected that the HTR would offer better thermal efficiency, better uranium utilisation, either with low enriched uranium fuel or high enriched uranium thorium fuel, better inherent safety and lower unit power costs. In the event all these potential advantages were demonstrated to be in principle achievable. This view is still shared today. In fact Very High Temperature Reactors is one of the concepts retained for Generation IV. Projects on constructing Modular Pebble Bed Reactors are under way. Here all available Dragon Project Reports (DPR) - approximately 1000 - are collected in electronic form. An index points to the reports (PDF format); each table in the report is accessible in EXCEL format with the aim of facilitating access to the data. These reports describe the design, experiments and modelling carried out over a period of 17 years. 2 - Related or auxiliary information: IRPHE-HTR-ARCH-01, Archive of HTR Primary Documents NEA-1728/01. 3 - Software requirements: Acrobat Reader, Microsoft Word, HTML Browser required

  9. Data handling at EBR-II [Experimental Breeder Reactor II] for advanced diagnostics and control work

    International Nuclear Information System (INIS)

    Lindsay, R.W.; Schorzman, L.W.

    1988-01-01

    Improved control and diagnostics systems are being developed for nuclear and other applications. The Experimental Breeder Reactor II (EBR-II) Division of Argonne National Laboratory has embarked on a project to upgrade the EBR-II control and data handling systems. The nature of the work at EBR-II requires that reactor plant data be readily available for experimenters, and that the plant control systems be flexible to accommodate testing and development needs. In addition, operational concerns require that improved operator interfaces and computerized diagnostics be included in the reactor plant control system. The EBR-II systems have been upgraded to incorporate new data handling computers, new digital plant process controllers, and new displays and diagnostics are being developed and tested for permanent use. In addition, improved engineering surveillance will be possible with the new systems

  10. NORA project offers unique reactor research and advanced training opportunities

    International Nuclear Information System (INIS)

    1961-01-01

    An international program for reactor research and advanced training for a period of three years has been established in connection with the Norwegian critical assembly NORA. The aim of the project is to determine, through integral experiments, the basic reactor physics data for lattices moderated with light-water, heavy-water or mixtures of heavy and light water, with fuels of different sizes and spacing, three different enrichments and compositions. The objectives, programme, and facilities are described in details

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-29

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

  12. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  13. Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Majumdar, S.

    1991-01-01

    Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

  14. An Overview of the International Reactor Physics Experiment Evaluation Project

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  15. Halden Reactor Project Workshop: Understanding Advanced Instrumentation and Controls Issues

    International Nuclear Information System (INIS)

    Beltracchi, L.

    1991-01-01

    A Halden Reactor Project Workshop on 'Understanding Advanced Instrumentation and Controls Issues' was held in Halden, Norway, during June 17-18, 1991. The objectives of the workshop were to (1) identify and prioritize the types of technical information that the Halden Project can produce to facilitate the development of man-machine interface guidelines and (2) to identify methods to effectively integrate and disseminate this information to signatory organizations. As a member of the Halden Reactor Project, the Nuclear Regulatory Commission (NRC) requested the workshop. This request resulted from the NRC's need for human factors guidelines for the evaluation of advanced instrumentation and controls. The Halden Reactor Project is a cooperative agreement among several countries belonging to the Organization for Economic Cooperation and Development (OECD). The US began its association with the Halden Project in 1958 through the Atomic Energy Commission. The project's activities are centered at the Halden heavy-water reactor and its associated man-machine laboratory in Halden, Norway. The research program conducted at Halden consists of studies on fuel performance and computer-based man-machine interfaces

  16. 20. Annual report. OECD Halden reactor project. 1979

    International Nuclear Information System (INIS)

    1981-01-01

    This is the Twentieth Annual Report on the OECD Halden Reactor Project, describing activities during 1979, the first year of the 1979-1981 Halden Agreement. Research work at the project is focussed on three areas: 1) In-core behaviour of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2) Prediction, surveillance and control of fuel and core performance, for which models of fuel and core behaviour are developed. 3) Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

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

    International Nuclear Information System (INIS)

    Saito, Ryusei; Kashihara, Shin-ichiro; Itoh, Shin-ichi

    1987-08-01

    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)

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

    International Nuclear Information System (INIS)

    1988-05-01

    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)

  19. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  20. Role of Halden Reactor Project for world-wide nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    McGrath, M.A.; Volkov, B.

    2011-07-01

    The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

  1. Role of Halden Reactor Project for world-wide nuclear energy development

    International Nuclear Information System (INIS)

    McGrath, M.A.; Volkov, B.

    2011-01-01

    The great interest for utilization of nuclear materials to produce energy in the middle of last century needed special investigations using first class research facilities. Common problems in the area of nuclear fuel development motivated the establishment of joint research efforts. The OECD Halden Reactor Project (HRP) is a good example of such a cooperative research effort, which has been performing for more than 50 years. During that time, the Halden Reactor evolved from a prototype heavy water reactor envisaged as a power source for different applications to a research reactor that is able to simulate in-core conditions of modern commercial power reactors. The adaptability of the Halden Reactor enables the HRP to be an important international test facility for nuclear fuels and materials development. The long-term international cooperation is based on the flexible HRP organizational structure which also provides the continued success. [1,2] This paper gives a brief history of the Halden Reactor Project and its contribution to world-wide nuclear energy development. Recent expansion of the Project to the East and Asian countries may also assist and stimulate the development of a nuclear industry within these countries. The achievements of the HRP rely on the versatility of the research carried out in the reactor with reliable testing techniques and in-pile instrumentation. Diversification of scientific activity in the areas of development of alternative energy resources and man-machine technology also provide the HRP with a stable position as one of the leaders in the world scientific community. All of these aspects are described in this paper together with current experimental works, including the investigation of ULBA (Kazakhstan) production fuel in comparison with other world fuel suppliers, as well as other future and prospective plans of the Project.(Author)

  2. Conceptual design studies of experimental and demonstration fusion reactors

    International Nuclear Information System (INIS)

    1978-01-01

    Since 1973 the FINTOR Group has been involved in conceptual design studies of TOKAMAK-type fusion reactors to precede the construction of a prototype power reactor plant. FINTOR-1 was the first conceptual design aimed at investigating the main physics and engineering constraints on a minimum-size (both dimensions and thermal power) tokamak experimental reactor. The required plasma energy confinement time as evaluated by various power balance models was compared with the values resulting from different transport models. For the reference design, an energy confinement time ten times smaller than neoclassical was assumed. This also implied a rather high (thermally stable) working temperature (above 20 keV) for the reactor. Other relevant points of the design were: circular plasma cross section, single-null axisymmetric divertor; lithium breeder, stainless steel structures, helium coolant; modular blanket and shield structure; copper-stabilized, superconducting Nb-Ti toroidal field and divertor coils; vertical field and transformer coils inside the toroidal coils; vacuum-tight containment vessel. Solutions involving air and iron transformer cores were compared. These assumptions led to a minimum size reactor with a thermal power of about 100MW and rather large dimensions (major radius of about 9m) similar to those of full-scale power reactors considered in other conceptual studies. The FINTOR-1 analysis was completed by the end of 1976. In 1977 a conceptual design of a Demonstration Power Reactor Plant (FINTOR-D) was started. In this study the main working assumptions differing from those of FINTOR-1 are: non-circular plasma cross section; plasma confinement compatible with trapped ion instabilities; cold (gas) blanket sufficient for wall protection (no divertor); wall loading between 1-3MW/m 2 and thermal power of a few GW. (author)

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

    International Nuclear Information System (INIS)

    Tachikawa, K.; Iida, H.; Nishio, S.; Tone, T.; Aota, T.; Iwamoto, T.; Niikura, S.; Nishizawa, H.

    1984-01-01

    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

  4. The strategy of experimental power reactor licensing in Indonesia

    International Nuclear Information System (INIS)

    Moch Djoko Birmano

    2015-01-01

    Currently, BATAN has being planned to develop Experimental Power Reactor (EPR), that is the research nuclear reactor that can generate power (electricity or heat). The EPR is planned will be built in the National Center for Research of Science and Technology (Puspiptek) area at Serpong, South Tangerang, Banten Province, with the choice of reactor types is HTGR with the power size of 10 MWth. As stated in the Act No. 10 year 1997 on Nuclear Power, that every construction and operation of nuclear reactors and other nuclear installations and decommissioning of nuclear reactors required to have a permit. Furthermore, the its implementation arrangements is regulated in Government Regulation (GR) No. 2 year 2014 on Licensing of Nuclear Installations and Nuclear Material Utilization, which contains the requirements and procedures for the licensing process since site, construction, commissioning, operation, and decommissioning, it means licensing is implemented during the activity of construction, operation and decommissioning of NPPs.While, for the more detailed licensing arrangements available in the guidelines of BAPETEN Chairman Regulation (BCR). This study was conducted to understand the legal and institutional aspects, types and stages, and the licensing process of RDE, and identify licensing strategy so that timely as planned. Methodologies used include the literature study, consultation with experts in BAPETEN, discussions in the national seminar including FGD. (author)

  5. Experimental evaluation of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1984-10-01

    The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

  6. Project experimental house in Caceres; Proyecto vivienda experimental en Caceres

    Energy Technology Data Exchange (ETDEWEB)

    Gamero, E.; Mendez, L.

    2008-07-01

    The project envisages to develop and efficient methodology for the design and building of social housing in the region of Extremadura, with the aim to obtaining sustainable buildings, with a better energy behaviour, and the use of renewable energy sources, in addition to ensuring a substantial increase of the quality of construction. the project is based on the construction of two isolated houses which will have the same geographical orientation and equal structural system, so they both will be subject to the same climate conditions. One of the houses (pattern house) will be built following exactly the construction being used for the Special Housing Plan, whilst the second on (Experimental house) will serve as a laboratory where to test and demonstrate different sustainability measures and renewable energy sources. (Author)

  7. Design project of the dosimetry control system in the independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. V; Album V: Predprojekat sistema dozimetrijske kontrole u nezavisnom kolu CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    Design project of the dosimetry control system in the independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels includes the following: calculations of CO{sub 2} gas activity, design of the dosimetry control system, review of the changes that should be done in the RA reactor building for installing the independent CO{sub 2} loop, specification of the materials with cost estimation, engineering drawings of the system. Predprojekat sistema dozimetrijske kontrole u nezavisnom kolu CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA sadrzi: proracun aktivnosti gasa CO{sub 2}, idejno resenje sistema dozimetrijske kontrole, pregled izmena koje bi trebalo izvrsiti u zgradi RA u vezi montaze nezavisnog kola CO{sub 2}, specifikaciju materijala sa cenom kostanja i graficku dokumentaciju predprojekta.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C.

    2011-01-01

    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)

  10. Experimental Investigation of Creep Behavior of Reactor Vessel Lower Head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1999-01-01

    The authors report a study which aimed at experimentally and numerically investigating and characterizing the failure of a reactor pressure vessel (RPV) lower head due to thermal and pressure loads generated by a severe accident. They present the experimental apparatus which is based on a scaled version of the lower part of a TMI-like reactor pressure vessel without vessel skirt. They report and comment the results obtained during the first five experiments: uniform heating and non penetrations, centre-peaked heat flux and no penetrations, edge-peaked heat flux and no penetrations, uniform heating with penetrations, edge-peaked heat flux with penetrations. They compare the third and fifth experience (those with edge-peaked heat flux)

  11. Experimental fuel channel for samples irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1984-12-01

    An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

  12. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

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

    International Nuclear Information System (INIS)

    Stefanova, S.; Chantoin, P.; Kolev, I.

    1994-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Chantoin, P; Kolev, I [eds.

    1994-12-31

    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.

  15. The Orphee reactor current status and proposed enhancement of experimental capabilities

    International Nuclear Information System (INIS)

    Breant, P.

    1990-01-01

    This report provides a description of the Orphee reactor, together with a rapid assessment of its experimental and research capabilities. The plans for enhancing the reactor's experimental capabilities are also presented. (author)

  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 studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

    1990-01-01

    Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

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

    International Nuclear Information System (INIS)

    1979-11-01

    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)

  19. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    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

  20. Halden Reactor Project activities, achievements and international collaboration

    International Nuclear Information System (INIS)

    Wiesenack, W.

    2003-01-01

    This paper concentrates on the Halden Project research programme related to fuel testing. An overview of ongoing tests on WWER fuel performance is also included. The ongoing and planned experiments containing WWER-related fuels and materials - Irradiation of Standard and Modified WWER Fuel (IFA-503) and Corrosion Testing of Different Cladding Alloys (IFA-638) - are presented. The future experiments involving WWER fuel and cladding types foreseen in of the Halden Reactor Project programme are given

  1. Achievements of European projects on membrane reactor for hydrogen production

    NARCIS (Netherlands)

    di Marcoberardino, G.; Binotti, M.; Manzolini, G.; Viviente, J.L.; Arratibel Plazaola, A.; Roses, L.; Gallucci, F.

    2017-01-01

    Membrane reactors for hydrogen production can increase both the hydrogen production efficiency at small scale and the electric efficiency in micro-cogeneration systems when coupled with Polymeric Electrolyte Membrane fuel cells. This paper discusses the achievements of three European projects

  2. Internal structure of reactor building for Madras Atomic Power Project

    International Nuclear Information System (INIS)

    Pandit, D.P.

    1975-01-01

    The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

  3. 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.

  4. Compilation of reactor-physical data of the AVR experimental reactor for 1982

    International Nuclear Information System (INIS)

    Werner, H.; Wawrzik, U.; Grotkamp, T.; Buettgen, I.

    1983-12-01

    Since the end of 1981 the calculation model AVR-80 has been taken as a basis for compiling reactor-physical data of the AVR experimental reactor. A brief outline of the operation history of 1982 is given, including the beginning of a large-scale experiment dealing with change-over from high enriched uranium to low enriched uranium. Calculations relative to spectral shift, diffusion, temperature, burnup, and recirculation of the fuel elements are described in brief. The essential results of neutron-physical and thermodynamic calculations and the characteristical data of the various types of fuel used are shown in tables and illustrations. (RF) [de

  5. ADVANTAGES, DISADVANTAGES, AND LESSONS LEARNED FROM MULTI-REACTOR DECOMMISSIONING PROJECTS

    International Nuclear Information System (INIS)

    Morton, M.R.; Nielson, R.R.; Trevino, R.A.

    2003-01-01

    This paper discusses the Reactor Interim Safe Storage (ISS) Project within the decommissioning projects at the Hanford Site and reviews the lessons learned from performing four large reactor decommissioning projects sequentially. The advantages and disadvantages of this multi-reactor decommissioning project are highlighted

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

    International Nuclear Information System (INIS)

    McDermott, M. D.; Griffin, C. D.; Michelbacher, J. A.; Earle, O. K.

    2000-01-01

    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

  7. Reactor-specific spent fuel discharge projections, 1984 to 2020

    International Nuclear Information System (INIS)

    Heeb, C.M.; Libby, R.A.; Holter, G.M.

    1985-04-01

    The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs

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

    International Nuclear Information System (INIS)

    Henning, C.D.; Gilleland, J.R.

    1988-01-01

    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

  9. Venus project : experimentation at ENEA's pilot site

    International Nuclear Information System (INIS)

    Bargellini, M.L.; Fontana, F.; Niccolai, L.; Scavino, G.; Mancini, R.; Levialdi, S.

    1996-12-01

    The document describes the ENEA's (Italian Agency for New Technologies, Energy and the Environment) experience in the Venus Project (Esprit III 6398). Venus is an advanced visual interface based on icon representation that permits to end-user to inquiry databases. VENUS interfaces to ENEA's databases: cometa materials Module, Cometa Laboratories Module and European Programs. This report contents the results of the experimentation and of the validation carried out in ENEA's related to the Venus generations. Moreover, the description of the architecture, the user requirements syntesis and the validation methodology of the VENUS systems have been included

  10. A study of reactor neutrino monitoring at the experimental fast reactor JOYO

    International Nuclear Information System (INIS)

    Furuta, H.; Fukuda, Y.; 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.

    2012-01-01

    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 24.3 m from the JOYO reactor core of 140 MW thermal power. The measured neutrino event rate from reactor on-off comparison was 1.11±1.24(stat.)±0.46(syst.) events/day. Although the statistical significance of the measurement was not enough, backgrounds in such a compact detector at the ground level were studied in detail and MC simulations were found to describe the data well. A study for improvement of the detector for future such experiments is also shown.

  11. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

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

    International Nuclear Information System (INIS)

    1986-11-01

    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/M 2 . 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

  13. Real-time numerical simulation with high efficiency for an experimental reactor system

    International Nuclear Information System (INIS)

    Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

    2006-01-01

    The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)

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

    International Nuclear Information System (INIS)

    Karpeta, C.

    1979-01-01

    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

  15. Project margins of advanced reactor design WWER-500

    International Nuclear Information System (INIS)

    Rogov, M.F.; Birukov, G.I.; Ershov, V.G.; Volkov, B.E.

    1994-01-01

    Project criteria for design of advanced WWER-500 reactor within design conditions are compared to the requirements of the Russian regulatory guides. Normal operation limits, safe operation limits for main anticipated operational occurrences and design limits accepted for design basis accidents are considered as in preliminary safety report. It is shown that the basic design criteria in the design of WWER-500 for the anticipated operational occurrences and for design basis accidents are more severe than required in the following regulatory guides General Safety Regulations for Nuclear Power Plants and Nuclear Safety Rules for Reactors of Nuclear Power Plants. This provides certain margins from safety point of view

  16. Research in nuclear reactor theory and experimental reactors; Istrazivanja u teoriji nuklearnih reaktora i ekspeimentalni reaktori

    Energy Technology Data Exchange (ETDEWEB)

    Pop-Jordanov, J [Elektrotehnicki fakultet, Beograd (Yugoslavia)

    1978-05-15

    The paper is devoted to the possibilities of using experimental reactors for scientific research in nuclear power with a stress on problems in nuclear reactor theory. The stationary and nonstationary neutron fields, burnup prediction and analyses as well as fuel element development and the corresponding role of test-reactors were dealt with. It was shown that the investigations in nuclear reactor theory in Yugoslavia were developing continuously and in a useful interaction with experiments on research reactors. The needs for continuing the work on fundamental problems in neutron transport theory and on improving the calculation methods for thermal power reactors, together with the improvement of performances of existing research systems, were pointed out. A new quality in scientific work could be obtained dealing with the problems connected to a possible introduction of test-reactors, and fast systems later on. It was also pleaded for the corresponding orientations in fundamental sciences. (author) Rad je posvecen mogucnostima koriscenja eksperimentalnih reaktora za naucna istrazivanja u nuklearnoj energetici, sa akcentom na probleme teorije nuklearnih reaktora. Obradjena su stacionarna i nestacionarna neutronska polja, predikcija i analize sagorevanja, kao i razvoj gorivnih elemenata te uloga test-reaktora u osvajanju njihove tehnologije. Pokazano je da su se istrazivanja u teoriji nuklearnih reaktora u nas odvijala kontinualno i u korisnoj interakciji sa eksperimentima na istrazivackim reaktorima. Istaknuta je potreba nastavljanja rada na fundamentalnim problemima transportne teorije neutrona i na usavrsavanju metoda proracuna termalnih enerrgetskih reaktora, uz poboljsanje performansi postojecih istrazivackih sistema. Novi kvalitet u naucnom radu bi predstavljala orijentacija na probleme vezane sa eventualnim uvodjenjem test-reaktora, a zatim i brzih sistema. Pledirano je i za odgovarajuca usmeravanja u fundamentalnim naukama. (author)

  17. Overview of the OECD-Halden reactor project

    International Nuclear Information System (INIS)

    Vitanza, Carlo

    2001-01-01

    The OECD Halden Reactor Project is an international network dedicated to enhanced safety and reliability of nuclear power plants. The Project operates under the auspices of the OECD Nuclear Energy Agency and aims at addressing and resolving issues relevant to safety as they emerge in the nuclear community. This paper gives a concise presentation of the Project goals and of its technical infrastructure. The paper contains also a brief overview of results from the programme carried out in the time period 1997-1999 and of the main issues contemplated for the 3-year programme period 2000-2002

  18. Magnet systems for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Miller, J.R.

    1988-01-01

    The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

  19. Manipulator system for remote maintenance of fusion experimental reactor

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Munakata, Tadashi; Murakami, Shin; Kondoh, Mitsunori.

    1991-01-01

    We have completed the conceptual design for a rail-mounted vehicle type remote maintenance system for the fusion experimental reactor (FER), which will be the first D-T burning reactor in Japan. We have fabricated a 1/5-scale model and confirmed the feasibility of the design. In this system, a rail is deployed into the vessel and supported at four horizontal ports. A vehicle then moves along the rail and handles in-vessel components with manipulators. The advantages of this concept are the high stiffness and high reliability of the rail, and the high mobility of the vehicle for efficient maintenance operations. In the FER, this concept is considered to be the first option for in-vessel maintenance. This paper describes the conceptual design of the system and the feasibility study using the 1/5-scale model. (author)

  20. Metal fires and their implications for advanced reactors. Part 3: Experimental and modeling results

    International Nuclear Information System (INIS)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-01-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  1. Research projects at the TRIGA-reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Buchberger, T.; Buchtela, K.; Hammer, J.; Miksovsky, A.; Veider, A.; Weber, H.W.; Zugarek, G.

    1986-01-01

    In 1985 the thermalizing column was modified to a beam tube with a conical collimator for neutron radiography. A highly sophisticated sample and cassette changer will be constructed in the next months. The central channel of the thermal column is also used for neutron radiography especially for small objects. The four beam tubes of the TRIGA-reactor are intensively used for neutron spectroscopy, small angle scattering, neutron interferometry and investigations of magnetic structures with polarized neutrons. The neutron activation installation in the piecing beam tube is permanently used for various sample analysis using a ultrafast pneumatic transfer system. In addition to these experiments directly related to the TRIGA-reactor other research projects are carried out, some of them under an IAEA research contract which are mostly focused towards nuclear safeguards such as the magnetic scanning of power reactor fuel assemblies or the laser surveillance system of spent fuel pools. (author)

  2. Experimental investigation of the MSFR molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

    In the paper experimental modelling and investigation of the MSFR concept will be presented. MSFR is a homogeneous, single region liquid fuelled fast reactor concept. In case of molten salt reactors 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 core is a single region homogeneous volume without internal structures, it is a difficult task to ensure stable flow field, which is strongly coupled to the volumetric heat generation. These considerations suggest that experimental modelling would greatly help to understand the flow phenomena in such geometry. A scaled and segmented experimental mock-up of MSFR was designed and built in order to carry out particle image velocimetry measurements. Basic flow behaviour inside the core region can be investigated and the measurement data can also provide resource for the validation of computational fluid dynamics models. Measurement results of steady state conditions will be presented and discussed.

  3. Remote welding and cutting techniques for fusion experimental reactors

    International Nuclear Information System (INIS)

    Onozuka, M.; Ishide, T.; Oda, Y.; Nagaoka, E.; Ue, K.; Kamei, H.

    1995-01-01

    Experimental investigation of the YAG laser cutting/welding and plasma gouging techniques has been conducted to examine their suitability for remote maintenance systems in future fusion experimental reactors. Using a hybrid beam coupling system, two laser beams of 500W and 740W powers were successfully combined to provide a 1,240W beam power. The combined laser was transmitted through the optical fiber for cutting and welding. The transmission loss for the beams is in the range of 13% to 14%, which is low. As for plasma gouging, the shallow gouging made a groove measuring 10 mm in width and 4 mm in depth on the stainless steel plates at a traversing speed of 75 cm/min, while the deep gouging made a groove of 12 mm in width and 7.5 mm in depth at a traversing speed of 50 cm/min. In addition, it was found that the shallow gouging did not leave byproducts from the material, providing a clean surface. Based on the findings, it is shown that the YAG laser cutting/welding and plasma gouging techniques can be us3ed for remote welding and cutting in future fusion experimental reactors

  4. Remote welding and cutting techniques for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Ishide, T.; Oda, Y.; Nagaoka, E.; Ue, K.; Kamei, H. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1995-12-31

    Experimental investigation of the YAG laser cutting/welding and plasma gouging techniques has been conducted to examine their suitability for remote maintenance systems in future fusion experimental reactors. Using a hybrid beam coupling system, two laser beams of 500W and 740W powers were successfully combined to provide a 1,240W beam power. The combined laser was transmitted through the optical fiber for cutting and welding. The transmission loss for the beams is in the range of 13% to 14%, which is low. As for plasma gouging, the shallow gouging made a groove measuring 10 mm in width and 4 mm in depth on the stainless steel plates at a traversing speed of 75 cm/min, while the deep gouging made a groove of 12 mm in width and 7.5 mm in depth at a traversing speed of 50 cm/min. In addition, it was found that the shallow gouging did not leave byproducts from the material, providing a clean surface. Based on the findings, it is shown that the YAG laser cutting/welding and plasma gouging techniques can be us3ed for remote welding and cutting in future fusion experimental reactors.

  5. Oscillation experiments techniques in CEA Minerve experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Antony, M.; Di-Salvo, J.; Pepino, A.; Bosq, J. C.; Bernard, D.; Leconte, P.; Hudelot, J. P.; Lyoussi, A. [CEA CADARACHE, DEN/DER/SPEx, 13108 Saint Paul-lez-Durance (France)

    2009-07-01

    This paper deals with experiments in the Minerve pool Zero Power Reactor. Minerve is mainly devoted to neutronics studies, in view to improve the calculation routes by reducing the uncertainties of the experimental databases for nuclides arising in plutonium and wastes management. Minerve experimental measurement programs are performed by using the oscillation technique. This experimental technique consists in a periodic insertion and extraction of samples containing the nuclide of interest in a well characterized neutron spectrum. The reactivity variation of the sample is compensated by a calibrated rotary automatic pilot using cadmium sectors. The normal accuracy for measurements of small-worth samples in Minerve by using such a technique is about 3% for absolute reactivity worth, including the uncertainties on the material balance and on the calibration step. Reactivity effects of less than 1.5 cent can be measured. The OSMOSE and the OCEAN programs have been carried out since 2005 and will last until 2011. These programs aim at improving, in different neutron spectra, the absorption cross sections of respectively a majority of the separated heavy nuclides from {sup 232}Th to {sup 245}Cm appearing during the reactor and the fuel cycle physics, and of current and future types of absorbers as Gd, Hf, Er, Dy and Eu. (authors)

  6. Project accent: graphite irradiated creep in a materials test reactor

    International Nuclear Information System (INIS)

    Brooking, M.

    2014-01-01

    Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)

  7. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    International Nuclear Information System (INIS)

    Johansen, R.

    2011-01-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  8. Radiation protection monitoring at the JOYO experimental fast reactor

    International Nuclear Information System (INIS)

    Ouchi, S.; Endo, K.; Susaki, T.

    1979-01-01

    This paper describes the radiation protection monitoring programme for the JOYO experimental fast reactor and some of the health physics problems experienced during the low-power nuclear tests. These include: a detailed description of the centralized radiation monitoring system; the methods and results of the individual monitoring systems; the results of operational monitoring for the handling of new plutonium fuel subassemblies; the evaluation of the external radiation dose rate around the primary coolant system; and the results of an experiment on the thermal dependence of some personnel dose meters. (author)

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

    International Nuclear Information System (INIS)

    Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

    1986-01-01

    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.%

  10. Experimental power reactor dc generator energy storage study

    International Nuclear Information System (INIS)

    Heck, F.M.; Smeltzer, G.S.; Myers, E.H.; Kilgore, L.

    1978-01-01

    This study covers the use of dc generators for meeting the Experimental Power Reactor Ohmic Heating Energy Storage Requirements. The dc generators satisfy these requirements which are the same as defined in WFPS-TME-038 which covered the use of ac generators and homopolar generators. The costs of the latter two systems have been revised to eliminate first-of-a-kind factors. The cost figures for dc generators indicate a need to develop larger machines in order to take advantage of the economy-of-scale that the large ac machines have. Each of the systems has its own favorable salient features on which to base a system selection

  11. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  12. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Cherepnin, Yu.S.; Bezzubtsev, V.S.; Gabaraev, B.A.

    2002-01-01

    Positive changes are currently taking place in nuclear power in the world. Power generation at Nuclear Power Plants (NPPs) is increasing and new units construction and completion rates are growing in some of leading countries. Considerable efforts are made for improving the safety of operating NPPs, effective use of nuclear fuel and solving the spent nuclear fuel and radioactive waste problems. Simultaneously, work are undertaken to develop new reactor technologies to reduce the fundamental drawbacks of conventional nuclear power, namely: insufficient safety, spent fuel and waste handling problems, nuclear material proliferation risk and poor economic competitiveness as compared to fossil-fuel energy sources. One the most important events in this field is an international project implemented by three agencies (OECD-IEA, OECD-NEA, IAEA) for comparative evaluation of new projects, development of Generation IV reactors underway in the US in cooperation with a number of Western countries and, finally, the initiative by Russian President V.V. Putin for consolidation the efforts of interested countries under auspices of IAEA to solve the problem of energy support for sustainable development of humankind, radical solution of non-proliferation problems and environmental sanitation of the Planet of Earth. The 44-th General Conference of IAEA in September 2000 supported the Initiative of Russian President and called all interested countries to unite efforts under the Agency's auspices in the International Project on Innovative Nuclear Reactors and Fuel Cycles to consider and select the most acceptable nuclear technologies of the 21-st century with regard for the drawbacks of today's nuclear power. Main objectivities of INPRO: Promotion of the availability of nuclear power for sustainable satisfaction of the energy needs in 21-st century; Consolidation of efforts by all interested INPRO participating countries (both owners and users of technologies) for joint development of

  13. Lessons learned in process control at the Halden Reactor Project

    International Nuclear Information System (INIS)

    Kennedy, W.G.

    1989-12-01

    This report provides a list of those findings particularly relevant to regulatory authorities that can be derived from the research and development activities in computerized process control conducted at the Halden Reactor Project. The report was prepared by a staff member of the US Nuclear Regulatory Commission working at Halden. It identifies those results that may be of use to regulatory organizations in three main areas: as support for new requirements, as part of regulatory evaluations of the acceptability of new methods and techniques, and in exploratory research and development of new approaches to improve operator performance. More than 200 findings arranged in nine major categories are presented. The findings were culled from Halden Reactor Project documents, which are listed in the report

  14. Actinide behavior in the Integral Fast Reactor. Final project report

    Energy Technology Data Exchange (ETDEWEB)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

  15. Actinide behavior in the Integral Fast Reactor. Final project report

    International Nuclear Information System (INIS)

    Courtney, J.C.

    1994-11-01

    The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ( 237 Np, 240 Pu, 241 Am, and 243 Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and weapons grade plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for seven day exposure in the Experimental Breeder Reactor-II which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction rates and neutron spectra. These experimental data increase the authors confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs

  16. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

    International Nuclear Information System (INIS)

    2017-01-01

    The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

  17. Experimental studies of tritium barrier concepts for fusion reactors

    International Nuclear Information System (INIS)

    Maroni, V.A.; Van Deventer, E.H.; Renner, T.A.; Pelto, R.H.; Wierdak, C.J.

    1976-01-01

    Ongoing experimental studies at ANL aimed at the development of methods to reduce tritium migration in fusion reactor systems currently include (1) work on the development of multilayered metal composites and impurity-coated refractory metals as barriers to tritium permeation in elevated temperature (greater than 300 0 C) structures and (2) investigations of the kinetics of tritium trapping reactions in inert gas purge streams under conditions that emulate fusion reactor environments. Significant results obtained thus far are (1) demonstration of greater than 50-fold reductions in the hydrogen permeability of stainless steel structures by using stainless steel-clad composites containing an intermediate layer of a selected copper alloy and (2) verification that surface-oxide coatings lead to greater than 100-fold reductions in the hydrogen permeability of vanadium, but that severe oxygen penetration and embrittlement of the vanadium occur at temperatures in the range from 300 to 800 0 C and under conditions of extremely low oxygen potential. Other considerations pertaining to the large-scale use of metal composites in fusion reactors are discussed, and progress in efforts to demonstrate the fabricability of metal composites is reviewed. Also presented are results of studies of the efficiencies of (1) CuO and CuO--MnO 2 beds in converting HT to HTO and (2) magnesium metal beds in converting HTO to HT

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

    International Nuclear Information System (INIS)

    Nakashima, Kunihiko; Yamamoto, Shin; Ohara, Yoshihiro; Watanabe, Kazuhiro; Mizuno, Makoto; Araki, Masanori; Uede, Taisei; Okano, Kunihiko.

    1987-09-01

    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)

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

    International Nuclear Information System (INIS)

    Drinovac, P.

    2006-01-01

    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.)

  20. Project management plan for Reactor 105-C Interim Safe Storage project

    International Nuclear Information System (INIS)

    Plagge, H.A.

    1996-09-01

    Reactor 105-C (located on the Hanford Site in Richland, Washington) will be placed into an interim safe storage condition such that (1) interim inspection can be limited to a 5-year frequency; (2) containment ensures that releases to the environmental are not credible under design basis conditions; and (3) final safe storage configuration shall not preclude or significantly increase the cost for any decommissioning alternatives for the reactor assembly.This project management plan establishes plans, organizational responsibilities, control systems, and procedures for managing the execution of Reactor 105-C interim safe storage activities to meet programmatic requirements within authorized funding and approved schedules

  1. Development step toward fusion power plant and role of experimental reactor ITER

    International Nuclear Information System (INIS)

    Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2005-01-01

    The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

  2. Programmes and projects for high-temperature reactor development

    International Nuclear Information System (INIS)

    Bogusch, Edgar; Hittner, Dominique

    2009-01-01

    An increasing attention has to be recognised worldwide on the development of High-Temperature Reactors (HTR) which has started in Germany and other countries in the 1970ies. While pebble bed reactors with spherical fuel elements have been developed and constructed in Germany, countries such as France, the US and Russia investigated HTR concepts with prismatic block-type fuel elements. The concept of a modular HTR formerly developed by Areva NP was an essential basis for the HTR-10 in China. A pebble bed HTR for electricity production is developed in South Africa. The construction is planned after the completion of the licensing procedure. Also the US is planning an HTR under the NGNP (Next Generation Nuclear Plant) Project. Due to the high temperature level of the helium coolant, the HTR can be used not only for electricity production but also for supply of process heat. Including its inherent safety features the HTR is an attractive candidate for heat supply to various types of plants e.g. for hydrogen production or coal liquefactions. The conceptual design of an HTR with prismatic fuel elements for the cogeneration of electricity and process heat has been developed by Areva NP. On the European scale the HTR development is promoted by the RAPHAEL (ReActor for Process heat, Hydrogen And ELectricity generation) project. RAPHAEL is an Integrated Project of the Euratom 6th Framework Programme for the development of technologies towards a Very High-Temperature Reactor (VHTR) for the production of electricity and heat. It is financed jointly by the European Commission and the partners of the HTR Technology Network (HTR-TN) and coordinated by Areva NP. The RAPHAEL project not only promotes HTR development but also the cooperation with other European projects such as the material programme EXTREMAT. Furthermore HTR technology is investigated in the frame of Generation IV International Forum (GIF). The development of a VHTR with helium temperatures above 900 C for the

  3. Latest feedback from a major reactor vessel dismantling project

    International Nuclear Information System (INIS)

    Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

    2015-01-01

    Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

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

    International Nuclear Information System (INIS)

    James, R.A.

    1996-01-01

    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)

  5. Japan: The Experimental Fast Reactor JOYO. Profile 12

    International Nuclear Information System (INIS)

    2017-01-01

    The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

  6. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  7. Experimental investigation of creep behavior of reactor vessel lower head

    International Nuclear Information System (INIS)

    Chu, T.Y.; Pilch, M.; Bentz, J.H.; Behbahani, A.

    1998-03-01

    The objective of the USNRC supported Lower Head Failure (LHF) Experiment Program at Sandia National Laboratories is to experimentally investigate and characterize the failure of the reactor pressure vessel (RPV) lower head due to the thermal and pressure loads of a severe accident. The experimental program is complemented by a modeling program focused on the development of a constitutive formulation for use in standard finite element structure mechanics codes. The problem is of importance because: lower head failure defines the initial conditions of all ex-vessel events; the inability of state-of-the-art models to simulate the result of the TMI-II accident (Stickler, et al. 1993); and TMI-II results suggest the possibility of in-vessel cooling, and creep deformation may be a precursor to water ingression leading to in-vessel cooling

  8. 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.

  9. The human factors and the safety of experimentation reactors

    International Nuclear Information System (INIS)

    Jeffroy, F.; Delaporte-Normier, M.L.

    2007-01-01

    Inside IRSN (Institute for Radiological protection and Nuclear Safety), the mission of the Human Factors Group is to assess the way operators of nuclear installations take into account the risks related to human activities. In the last few years, IRSN has been involved in the safety analysis of different installations where Cea develops research programs, in particular experimental reactors. The first part of this article presents the methodology used by IRSN to evaluate how operators take into account risks related to human activities. This methodology is made up of 4 steps: 1) the identification of the human activities that convey a risk for the installation nuclear safety (safety-sensitive activities), for instance in the case of the Masurca reactor, it has been shown that errors made during the manufacturing of fuel tubes can lead to a criticality accident; 2) listing all the dispositions or arrangements taken to make human safety-sensitive activities more reliable; 3) checking the efficiency of such dispositions or arrangements; and 4) assessing the ability of the operators to generate the adequate dispositions or arrangements. The second part highlights the necessity to develop inside these research installations an organisation that facilitates cooperation between experimenters and operators

  10. Neutron thermalization in reactor lattice cells: An NPY-project report

    International Nuclear Information System (INIS)

    Stamm'ler, R.J.J.; Takac, S.M.; Weiss, Z.J.

    1966-01-01

    The NPY-Project is a joint research programme in reactor physics between Norway, Poland, Yugoslavia and the International Atomic Energy Agency. One of the tasks of the project was to make a theoretical and experimental investigation of the phenomena of neutron thermalization in lattice cells, and this work is covered by the present monograph. The different lattices of the zero-power assemblies in the NPY countries offered ample opportunity for the theoreticians and experimentalists to test and compare their methods, and the exchange of experiences was stimulating and valuable. 85 refs, 26 figs, 19 tabs

  11. Primary system thermal-hydraulic simulation of a experimental pool type research fast reactor

    International Nuclear Information System (INIS)

    Borges, E.M.; Braz Filho, F.A.

    1993-01-01

    The first step of the Fast Reactor Program (REARA) is the design of an experimental reactor. To this end a 5 MW t pool type reactor was adapted. The objective of this work is to evaluate the reactor behaviour at the on set protected accidents. The program NALAP was used in this study and the results showed the outstanding safety margins that this reactor type presents inherently. (author)

  12. European lead fast reactor (ELSY and LEADER projects)

    International Nuclear Information System (INIS)

    Alemberti, Alessandro; Carlsson, Johan; Malambu, Edouard; Orden, Alfredo; Cinotti, Luciano; Struwe, Dankward; Agostini, Pietro; Monti, Stefano

    2010-01-01

    The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY project. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative DHR systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been

  13. International project GT-MHR - New generation of nuclear reactors

    International Nuclear Information System (INIS)

    Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

    2001-01-01

    Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

  14. International Thermonuclear Experimental Reactor: Physics issues, capabilities and physics program plans

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1997-01-01

    Present status and understanding of the principal plasma-performance determining physics issues that affect the physics design and operational capabilities of the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2 (International Atomic Energy Agency, Vienna, 1994)] are presented. Emphasis is placed on the five major physics-basis issues emdash energy confinement, beta limit, density limit, impurity dilution and radiation loss, and the feasibility of obtaining partial-detached divertor operation emdash that directly affect projections of ITER fusion power and burn duration performance. A summary of these projections is presented and the effect of uncertainties in the physics-basis issues is examined. ITER capabilities for experimental flexibility and plasma-performance optimization are also described, and how these capabilities may enter into the ITER physics program plan is discussed. copyright 1997 American Institute of Physics

  15. Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

    The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors. (author)

  16. The importance of project networking for the replacement research reactor

    International Nuclear Information System (INIS)

    Whitbourn, G.

    2003-01-01

    When the HIFAR research reactor was commissioned in 1958 it was both constructed and regulated by the then Australian Atomic Energy Commission. The situation now is much more complicated, with an independent regulator, The Australian Radiation Protection and Nuclear Safety Agency (ARPANSA) and oversight by national security agencies and the Australian Safeguards and Non proliferation Organisation (ASNO). In July 2000 ANSTO contracted INVAP SE a suitably qualified and experienced nuclear organisation based in Argentina to provide the Replacement Research Reactor (RRR). INVAP subcontracted an Australian entity, a joint venture between John Holland and Evans Deakin Industries (JHEDI) to provide resources in Australia. There is an international network of over 100 subcontractors providing services, products and materials to INVAP and JHEDI and a significant number of contractors providing project support services to ANSTO. The interaction of all these entities to provide the RRR is a significant networking challenge, involving a complex network of legal, contractual and functional relationships and communication processes

  17. 105-C Reactor interim safe storage project technology integration plan

    International Nuclear Information System (INIS)

    Pulsford, S.K.

    1997-01-01

    The 105-C Reactor Interim Safe Storage Project Technology Integration Plan involves the decontamination, dismantlement, and interim safe storage of a surplus production reactor. A major goal is to identify and demonstrate new and innovative D and D technologies that will reduce costs, shorten schedules, enhance safety, and have the potential for general use across the RL complex. Innovative technologies are to be demonstrated in the following areas: Characterization; Decontamination; Waste Disposition; Dismantlement, Segmentation, and Demolition; Facility Stabilization; and Health and Safety. The evaluation and ranking of innovative technologies has been completed. Demonstrations will be selected from the ranked technologies according to priority. The contractor team members will review and evaluate the demonstration performances and make final recommendations to DOE

  18. The international thermonuclear experimental reactor and the future of nuclear fusion energy

    International Nuclear Information System (INIS)

    Pan Chuanhong

    2010-01-01

    Energy shortage and environmental problems are now the two largest challenges for human beings. Magnetic confinement nuclear fusion, which has achieved great progress since the 1990's, is anticipated to be a way to realize an ideal source of energy in the future because of its abundance, environmental compatibility, and zero carbon release. Exemplified by the construction of the International Thermonuclear Experimental Reactor (ITER), the development of nuclear fusion energy is now in its engineering phase, and should be realized by the middle of this century if all objectives of the ITER project are met. (author)

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

    International Nuclear Information System (INIS)

    Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

    2013-01-01

    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)

  20. Experimental measurements in the BYU controlled profile reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tree, D.R.; Black, D.l.; Rigby, J.R.; McQuay, M.Q.; Webb, B.W. [Brigham Young University, Provo, UT (United States). Dept. of Mechanical Engineering

    1998-09-01

    Over the past decade the Controlled Profile Reactor (CPR) has been used to obtain extensive combustion data sets. CPR is a small scale (0.2-0.4 MW) combustion facility that has been used to obtain data for model validation, the testing of new combustion concepts, and the development of new combustion instruments. This review of the past ten years of research completed in the CPR includes a description of the reactor and instrumentation used, a summary of three experimental data sets which have been obtained in the reactor, and a description of novel tests and instrumentation. Measurements obtained include gas species, gas temperature, particle velocity, particle size, particle number density, particle-cloud temperature profiles, radiation and total heat flux to the wall, and wall temperatures. Species data include the measurement of CO, CO{sub 2}, NO, NO{sub x}, O{sub 2}, NH{sub 3} and HCN. The three combustion studies included one with natural gas combustion in a swirling flow, and two pulverized-coal combustion studies involving Utah Blind Canyon and Pittsburgh No. 8 coals. Most, but not all of the above measurements were obtained in each study. The second coal study involving the Pittsburgh No. 8 coal contained the most complete set of data and is described in detail. Novel combustion instrumentation includes the use of Coherent Anti-Stokes Raman Spectroscopy (CARS) to measure gas temperature. Novel combustion experiments include the measurement of NO{sub x} and burnout with coal-char blends. The measurements have led to an improved understanding of the combustion process and an understanding of the strengths and weaknesses associated with different aspects of comprehensive combustion models. 67 refs., 26 figs., 9 tabs.

  1. European supercritical water cooled reactor (HPLWR Phase 2 project)

    International Nuclear Information System (INIS)

    Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

    2010-01-01

    The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

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

    International Nuclear Information System (INIS)

    Nakashima, Kunihiko; Okano, Kunihiko; Miyamoto, Kazuhiro.

    1987-09-01

    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)

  3. Oak Ridge Tokamak experimental power reactor study scoping report

    International Nuclear Information System (INIS)

    Roberts, M.

    1977-03-01

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

  4. The European Fast Breeder Reactor project (EFR). Should Germany withdraw from the project?

    International Nuclear Information System (INIS)

    1993-01-01

    The Fed. Min. of Research and Technology has discontinued its financial aid to industry for the development of fast breeder reactor technology at the end of 1991, and a termination of the EFR project is in sight. There is, however, no need for the German government to withdraw from the EFR agreement concluded in 1984. (orig.) [de

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

    International Nuclear Information System (INIS)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

    2003-01-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10 14 ncm -2 s -1 and a fast flux of 6,4.10 14 ncm -2 s -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)

  6. General safety orientations of the Jules Horowitz Reactor Project (JHRP)

    International Nuclear Information System (INIS)

    Tremodeux, P.; Fiorini, G.L.

    2000-01-01

    After a brief reminder of the JHR purpose, the document outlines the General Safety related Orientations/Recommendations used for the design and the safety assessment of the facility. As far as the JHR design is new, the safety philosophy adopted for this reactor will be as consistent as possible with that recommended for future (power...) reactors. The general recommendations developed in the paper are: the general nuclear safety approach for the design, operation and analysis with, in particular, the adoption of the Defence In Depth principle; the general safety objectives in terms of radiological consequences; the use of Probabilistic Safety Studies; quality assurance. The 'Defence in Depth' concept using amongst others the 'Barrier' principle remains the basis of the JHR safety. 'Defence In Depth' is applied both to design and operation. Its adequacy is checked during the safety assessment and the paper gives the technical recommendations that should allow the designer to implement this concept into the final design. Built mainly for experimental irradiation the JHR facilities will be handled according to conventional or new operation rules which could put materials under stress and entail handling errors. Specific recommendations are defined to take into account the corresponding peculiarities; they are discussed in the paper. The safety design of the JHR takes into account the experience accumulated through the CEA experimental irradiation programmes, which represents several dozen reactor years; the consultation of CEA reactor facilities operators is ongoing. The corresponding feedback is shortly described. Recommendations related to maintenance and associated operation are indicated as well as those regarding the human factor. Details are given on the JHR safety practical implementation through the CEA/DRN Safety approach. Details of the corresponding Safety Objectives are also discussed. Finally the designer position on the role of probabilistic safety

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

    International Nuclear Information System (INIS)

    Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu.

    1987-08-01

    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)

  8. Superconducting coil design for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Turner, L.R.; Wang, S.T.; Smelser, P.

    1977-01-01

    Superconducting toroidal field (TF) and polodial-field (PF) coils have been designed for the proposed Argonne National Laboratory experimental power reactor (EPR). Features of the design include: (1) Peak field of 8 T at 4.2 K or 10 T at 3.0 K. (2) Constant-tension shape for the TF coils, corrected for the finite number (16) of coils. (3) Analysis of errors in coil alignment. (4) Comparison of safety aspects of series-connected and parallel-connected coils. (5) A 60 kA sheet conductor of NbTi with copper stabilizer and stainless steel for support. (6) Superconducting PF coils outside the TF coils. (7) The TF coils shielded from pulsed fields by high-purity aluminum

  9. IAEA Assistance in the development of new research reactor projects

    Energy Technology Data Exchange (ETDEWEB)

    Borio di Tigliole, Andrea; Bradley, Ed; Zhukova, Anastasia; Adelfang, Pablo [International Atomic Energy Agency, Research Reactor Section, Vienna (Austria); Shokr, Amgad [International Atomic Energy Agency, Research Reactor Safety Section, Vienna (Austria); Ridikas, Danas [International Atomic Energy Agency, Physics Section, Vienna (Austria)

    2015-08-15

    A research reactor (RR) project is a major undertaking that requires careful preparation, planning, implementation and investment in time, money, and human resources. In recent years, the interest of IAEA Member States in developing RR programmes has grown significantly, and currently, several Member States are in different stages of new RR projects. The majority of these countries are building their first RR as a key national facility for the development of their nuclear science and technology programmes, including nuclear power. In order to support Member States in such efforts, the IAEA in 2012 published the Nuclear Energy Series Report No. NP-T-5.1 on Specific Considerations and Milestones for a Research Reactor Project. To provide further support, the IAEA also published a document to assist Member States in the preparation of the bid invitation specification for the purchase of a RR. The IAEA will also continue to provide assistance for human resources development of the Member States establishing their first RR, and to facilitate sharing experience and knowledge among Member States through its programmatic activities including expert mission services, technical meetings, training courses and workshops addressing relevant technical and safety topics. This paper presents the IAEA assistance and services provided to the Member States considering new RRs, with particular emphasis on those establishing their first RR, including elaboration on the services mentioned above.

  10. Operational report, basics of the preliminary project for reconstruction of the RA reactor ventilation system

    International Nuclear Information System (INIS)

    Martinc, R.

    1982-01-01

    This report presents the summary of results needed for designing the reconstruction of the RA reactor ventilation system. The final complete report will published later, including results of possible additional analyses according to the needs of main detailed project. Due to the size of the facility the need of filters for particulates is expected in systems V-1 and V-2, as well as revision of the system V-4. Adsorption filters are not needed in case of operation under regular working conditions. It would be favorable, but not indispensable to construct an adsorption filter that could be used occasionally at the input of V-2 system under the upper water shielding. This filter would be switched on when dehermetization of the reactor is planned and increased radioactivity is indicated in the gas system or experimental space of the RA reactor as well as in case of fuel elements failure. The aim of applying filters is shortening the time these elements would spend in the hermetized reactor i.e. their fast removal from the reactor. It is indispensable to build-in 'emergency' highly efficient adsorption filters in the V-2 system which would be switched on in case of accidents that would cause fuel elements meltdown (with simultaneous accident dehermetization of the heavy water system) [sr

  11. The reprocessing of fast reactor fuels - the TOR project

    International Nuclear Information System (INIS)

    Calame-Longjean, A.; Le Bouhellec, J.; Schwob, Y.

    1982-01-01

    A description is given of development work on the proposed new French facility for the reprocessing of fast reactor fuel. This is the TOR facility (Traitement des Oxydes Rapides). Block diagrams give details of the TOR project as a whole and of the main line and R and D line of the TOR 1 facility which is a new works devoted to the head of the process. Modifications to existing plant which will form the TOR 2 and TOR 3 facilities are also described. (U.K.)

  12. A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodiac, F.; Hudelot, JP.; Lecerf, J.; Garnier, Y.; Ritter, G. [CEA, DEN, CAD/DER/SRES/LPRE, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Gueton, O.; Colombier, AC. [CEA, DEN, CAD/DER/SPRC/LPN, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Domergue, C. [CEA, DEN, CAD/DER/SPEx/LDCI, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimental program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)

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

    NARCIS (Netherlands)

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

    1987-01-01

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

  14. Evaluation of nuclear facility decommissioning projects. Project summary report, Elk River Reactor

    International Nuclear Information System (INIS)

    Miller, R.L.; Adams, J.A.

    1982-12-01

    This report summarizes information concerning the decommissioning of the Elk River Reactor. Decommissioning data from available documents were input into a computerized data-handling system in a manner that permits specific information to be readily retrieved. The information is in a form that assists the Nuclear Regulatory Commission in its assessment of decommissioning alternatives and ALARA methods for future decommissionings projects. Samples of computer reports are included in the report. Decommissioning of other reactors, including NRC reference decommissioning studies, will be described in similar reports

  15. 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.

  16. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

    Mourogov, V. M.; Juhn, P. E.

    2003-01-01

    In response to two IAEA General Conference Resolutions in September 2000, the IAEA has launched the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) in May 2001. As of February 2003, 12 IAEA Member States and the European Commission have become members of INPRO. In total, 19 cost-free experts have been nominated by these Member States and the European Commission to work for the INPRO project at the IAEA. Four meetings of the INPRO Steering Committee (SC), which is the decision and review body of INPRO, were held, two in 2001 and another two in 2002. The objective of INPRO, which is composed of two phases (Phase 1 and Phase 2), is to support safe, economic and proliferation resistant use of nuclear technology, in a sustainable manner, to meet the global energy needs in the next 50 years and beyond. During Phase 1, work is also subdivided in two sub phases: The currently on-going Phase 1A is focussing on the selection of criteria and development of methodologies and guidelines for the comparison of different reactor and fuel cycle concepts and approaches, taking into account the compilation and review of such concepts and approaches, and determination of user requirements in the areas of economics; environment; safety; proliferation-resistance; and cross cutting issues. The preliminary results of Phase 1A with respect to user requirements are summarized in the paper

  17. Summary report of the experimental fast reactor JOYO MK-III performance test

    International Nuclear Information System (INIS)

    Maeda, Yukimoto; Aoyama, Takafumi; Yoshida, Akihiro

    2004-03-01

    An upgrading project (MK-III project) was started to improve the irradiation capability of the experimental fast reactor JOYO. In this project, core replacement and increase of the reactor thermal power by the factor 1.4 were necessary for increasing the maximum fast neutron flux by the factor 1.3 and doubling the capacity for irradiation rigs. The modification of the cooling system that included the replacement of the main intermediate heat exchangers and the dump heat exchangers was completed in September 2000. After a series of system function tests, the performance test, of which objective is to fully characterize the upgraded core and heat transfer system, was started in June 2003. Twenty eight tests were selected and carried out as performance test, in order to confirm that the whole plant satisfy the design criteria and have sufficient characteristics (data necessary for safe and steady operation, core management, reactor control and monitoring) as an irradiation bed. After attaining the initial criticality of the core on 2nd July 2003, core characteristics (the excess reactivity, the isotherm temperature reactivity coefficient, the power reactivity coefficient and so on), plant characteristics (the plant heat balance, the adjustment of the temperature control system, the plant behavior at transient), shielding characteristics (dose rate distribution). As the result, it was confirmed that all the criteria regulated was satisfied and the core and plant have sufficient margins for full power operation, which was increased by the factor 1.4. Especially, nuclear analysis accuracy was verified by comparing the calculation with measured core characteristics of the initial core which consists of fifty five fresh fuel subassemblies. The operational data which is supposed to be useful for developing in-core anomaly detection system were also obtained. The operation manual and training simulator and design of next reactor development were revised based on the results

  18. Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems

    International Nuclear Information System (INIS)

    Henning, C.D.; Miller, J.R.

    1988-01-01

    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 magnetic 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

  19. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report

    International Nuclear Information System (INIS)

    1985-01-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training

  20. Experimental analysis of flowrates distribution features in double-loop reactor channels

    International Nuclear Information System (INIS)

    Avdeev, E.F.; Chusov, I.A.

    2013-01-01

    Experimental data on the flowrate distribution in working channels dummies of a research reactor model with double-loop configuration are presented in the paper. The procedures of experiments and received experimental data processing are provided in details [ru

  1. Regulation for installation and operation of experimental-research reactor

    International Nuclear Information System (INIS)

    1979-01-01

    The ordinance is stipulated under the Law for regulation of nuclear raw materials, nuclear fuel materials and reactors and the provisions for installation and operation of reactor in the order for execution of the law. Basic concepts and terms are defined, such as, radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; preserved area; inspected surrounding area and employee. An application for permission of installation of reactor shall list such matters as: the maximum continuous thermal output of reactor; location and general construction of reactor facilities; construction and equipment of the main reactor and other facilities for nuclear fuel materials; cooling and controlling system and radioactive waste, etc. An operation plan of reactor for three years shall be filed till January 31 of the fiscal year preceding that one the operation begins. Records shall be made and kept for specified periods respectively on inspection of reactor facilities, operation, fuel assembly, radiation control, maintenance, accidents of reactor equipment and weather. Detailed rules are settled for entrance limitation to controlled area, exposure dose, inspection, check up and regular independent examination of reactor facilities, operation of reactor, transportation of substances contaminated by nuclear fuel materials within the works and storage, etc. (Okada, K.)

  2. 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)

    2012-12-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-10-01

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

  4. Evaluation guide for the international reactor physics experiments evaluation project (IRPhEP)

    International Nuclear Information System (INIS)

    Yamaji, Akifumi

    2013-01-01

    At present, there is an urgent need to preserve integral reactor physics experimental data including separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. The International Reactor Physics Evaluation Project (IRPhEP) was initiated as a pilot activity in 1999 by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. This document outlines the general presentation guidelines that evaluators should follow for the description of the experiments and all relevant experimental data in order to ensure the consistency between the evaluations published in the final Handbook. Publication templates will be used to ensure this consistency and will follow the general scheme below: 1 - Experiment identification number; 2- Date; 3 - Name of experiment (Purpose of experiment, Phenomena measured and scope); 4 - Name or designation of experimental programme; 5 - Description of facility; 6 - Description of test or experiment (Experimental configuration, Core life cycle, Experimental limitations or shortcomings); 7 - Phenomena measured (Description of results and analysis, Special features and characteristics of experiment, Measurement systems/methods and uncertainties); 8 - Duplicate or complementary experiments / other related experiments; 9 - Status of completion of the evaluation; 10 - References (pointer to evaluation, archive if available, otherwise generic bibliographic reference); 11 - Authors/ organisers 12 - Material available

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-16

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

  6. Development and experimental qualification of a calculation scheme for the evaluation of gamma heating in experimental reactors. Application to MARIA and Jules Horowitz (JHR) MTR Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tarchalski, M.; Pytel, K.; Wroblewska, M.; Marcinkowska, Z.; Boettcher, A.; Prokopowicz, R. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Sireta, P.; Gonnier, C.; Bignan, G. [CEA, DEN, Reactor Studies Department, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Fourmentel, D.; Barbot, L.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C.; Brun, J. [Aix Marseille Universite, CNRS, Universite de Toulon, IM2NP UMR 7334, 13397, Marseille (France); Jagielski, J. [NCBJ Institute, MARIA Reactor, ul.Andrzeja Soltana 7, 05-400 Swierk (Poland); Institute of Electronic Materials Technolgy, Wolczynska 133, 01-919 Warszawa (Poland); Luks, A. [Institute of Heat Engineering, Nowowiejska 21/25, 00-665 Warsaw (Poland)

    2015-07-01

    Precise computational determination of nuclear heating which consists predominantly of gamma heating (more than 80 %) is one of the challenges in material testing reactor exploitation. Due to sophisticated construction and conditions of experimental programs planned in JHR it became essential to use most accurate and precise gamma heating model. Before the JHR starts to operate, gamma heating evaluation methods need to be developed and qualified in other experimental reactor facilities. This is done inter alia using OSIRIS, MINERVE or EOLE research reactors in France. Furthermore, MARIA - Polish material testing reactor - has been chosen to contribute to the qualification of gamma heating calculation schemes/tools. This reactor has some characteristics close to those of JHR (beryllium usage, fuel element geometry). To evaluate gamma heating in JHR and MARIA reactors, both simulation tools and experimental program have been developed and performed. For gamma heating simulation, new calculation scheme and gamma heating model of MARIA have been carried out using TRIPOLI4 and APOLLO2 codes. Calculation outcome has been verified by comparison to experimental measurements in MARIA reactor. To have more precise calculation results, model of MARIA in TRIPOLI4 has been made using the whole geometry of the core. This has been done for the first time in the history of MARIA reactor and was complex due to cut cone shape of all its elements. Material composition of burnt fuel elements has been implemented from APOLLO2 calculations. An experiment for nuclear heating measurements and calculation verification has been done in September 2014. This involved neutron, photon and nuclear heating measurements at selected locations in MARIA reactor using in particular Rh SPND, Ag SPND, Ionization Chamber (all three from CEA), KAROLINA calorimeter (NCBJ) and Gamma Thermometer (CEA/SCK CEN). Measurements were done in forty points using four channels. Maximal nuclear heating evaluated from

  7. Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1993-02-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

  8. Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1992-01-01

    This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

  9. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

    1999-01-01

    The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

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

    International Nuclear Information System (INIS)

    Stekly, Z.J.J.; Lucas, E.J.

    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

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

    International Nuclear Information System (INIS)

    Nakashima, Kunihiko; Ishigaki, Yukio; Ozaki, Akira; Yamane, Minoru.

    1987-09-01

    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)

  12. Upgrade of the experimental facilities of the ORPHEE reactor

    International Nuclear Information System (INIS)

    Farnoux, B.; Breant, P.

    1993-01-01

    At the time of the design, the ORPHEE reactor has been equipped with a set of up-to-date experimental facilities such as nine tangential and horizontal beam holes, one hot source, two hydrogen cold sources and six neutron guides. After more than ten years of operations, all the neutron beams are now used by about twenty five spectrometers. A modernisation program is under progress with a two fold aim: upgrade of the existing facilities and creation of new beams. Some details of the six following points will be described: 1) replacement of the flat cold source cell by an hollow cylinder in order first to increase the cold neutron flux and secondly to facilitate the extraction of new cold neutron beams. 2) replacement of the old neutron guide elements coated with natural nickel by new elements with isotopic nickel or super mirror coating. 3) modification of the curvature of some existing neutron guides in order to increase the wavelength band transmission. 4) creation of new cold neutron beams by installation of benders on the existing neutron guides. 5) design of new cold neutron guides and a new guide hall. 6) design of a thermal neutron guide. The two last points will made extensive use of super mirrors allowed by new technical developments done at the Laboratoire LEON BRILLOUIN in connection with industry. (author)

  13. Beam heating requirements for a tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Bertoncini, P.J.; Brooks, J.N.; Fasolo, J.A.; Stacey, W.M. Jr.

    1976-01-01

    Typical beam heating requirements for effective tokamak experimental power reactor (TEPR) operation have been studied in connection with the Argonne preliminary conceptual TEPR design. For an ignition level plasma (approximately 100 MWt fusion power) for the nominal case envisioned, the neutral beam is only used to heat the plasma to ignition. This typically requires a beam power output of 40 MW at 180 keV for about 3 sec with a total energy of 114 MJ supplied to the plasma. The beam requirements for an ignition device are not very sensitive to changes in wall-sputtered impurity levels or plasma resistivity. For a plasma that must be driven due to poor confinement, the beam must remain on for most of the burn cycle. For representative cases, beam powers of approximately 23 MW are required for a total on-time of 20 to 50 sec. Reqirements on power level, beam energy, on-time, and beam-generation efficiency all represent considerable advances over present technology. For the Argonne TEPR design, a total of 16 to 32 beam injectors is envisioned. For a 40-MW, 180-keV, one-component beam, each injector supplies about 7 to 14 A of neutrals to the plasma. For positive ion sources, about 50 to 100 A of ions are required per injector and some form of particle and/or energy recycling appears to be essential in order to meet the power and efficiency requirements

  14. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  15. Tokamak experimental power reactor conceptual design. Volume I

    International Nuclear Information System (INIS)

    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 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

  16. The collaborative project on European sodium fast reactor (CP ESFR project)

    International Nuclear Information System (INIS)

    Fiorini, Gian-Luigi

    2010-01-01

    The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The paper also gives insights concerning the so called 'working horses' cores and systems which are provided by CEA and AREVA and that will be used as a basis to test the performances and assess the pertinence of innovative solutions. The CP ESFR merges the contribution of 25 European partners (EU + CH); it will be performed under the aegis of the 7th Euratom FP under the Area - Advanced Nuclear Systems with a refund from the European Commission. It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follows: an improved safety with in particular the achievement of a robust architecture vis-a-vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk similar to that of the other means of energy production; a flexible and robust management of nuclear materials and especially waste reduction through Minor Actinides burning

  17. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  18. Advanced nuclear reactor public opinion project. Interim report

    Energy Technology Data Exchange (ETDEWEB)

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  19. First (Kick-Off) Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on Sodium Properties and Safe Operation of Experimental Facilities in Support of the Development and Deployment of Sodium-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    General information on recent IAEA activities in the field of fast reactors, including on-going and planned FR-related CRPs, as well as a review of the NAPRO CRP content, objectives, participants, planning and expected outcomes/deliverables, was provided. It was reminded that the NAPRO CRP is being organized in 3 main Work Packages (WP): • WP 1: collection, expert assessment and dissemination of consistent Na property data such as: surface tension, saturation vapor pressure and emissivity, thermo-dynamic behavior of ternary oxides in Na, and solubility - diffusivity of metallic impurities, etc. • WP 2: development of guidelines and best practices for Na facility design and operation, including fill and drain, purification, out-gassing prior to filling, Na storage, component handling, drying of Na piping after repair, etc. • WP 3: development of guidelines and best practices for Na facility safety, including prevention and mitigation of Na leaks, prevention and detection of Na fires, assessment of Na impact in the environment after accidental release, hydrogen hazards in cleaning facilities, etc

  20. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  1. Repair/maintenance design for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1978-10-01

    Repair and maintenance design for JXFR has been studied. The reactor is in eight modules so that a damaged module alone can be separated from the other modules and transferred from the reactor room to a repair shop. Design work covers overhaul procedure, dismounting equipments (overhead cranes, auto welder/cutter and remote handling equipments), transport system of a module (module mounting carriages and rotating carriage), repair equipment for blanket, earthquake-proof analysis of the reactor, reactor room structure, repair shop layout, management of radioactive wastes, time and the number of persons required for overhaul etc. Though the repair and maintenance system is almost complete, there still remain problems for further study in joints of blanket cooling piping, auto welder/cutter and earthquake-proof strength in reactor disassemblage. More detailed studies and R and D are necessary for engineering perfection. (author)

  2. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  3. Numerical and experimental investigation of geometric parameters in projection welding

    DEFF Research Database (Denmark)

    Kristensen, Lars; Zhang, Wenqi; Bay, Niels

    2000-01-01

    parameters by numerical modeling and experimental studies. SORPAS, an FEM program for numerical modeling of resistance welding, is developed as a tool to help in the phase of product design and process optimization in both spot and projection welding. A systematic experimental investigation of projection...... on the numerical and experimental investigations of the geometric parameters in projection welding, guidelines for selection of the geometry and material combinations in product design are proposed. These will be useful and applicable to industry.......Resistance projection welding is widely used for joining of workpieces with almost any geometric combination. This makes standardization of projection welding impossible. In order to facilitate industrial applications of projection welding, systematic investigations are carried out on the geometric...

  4. Oak Ridge Tokamak experimental power reactor study reference design

    International Nuclear Information System (INIS)

    Roberts, M.; Bettis, E.S.

    1975-11-01

    A Tokamak EPR Reference Design is presented as a basis for further design study leading to a Conceptual Design. The set of basic plasma parameters selected--minor radius of 2.25 m, major radius of 6.75 m, magnetic field on axis of 4.8 T and plasma current of 7.2 MA--should produce a reactor-grade plasma with a significant neutron flux, even with the great uncertainty in plasma physics scaling from present experience to large sizes. Neutronics and heat transfer calculations coupled with mechanical design and materials considerations were used to develop a blanket and shield capable of operating at high temperature, protecting the surrounding coils, being maintained remotely and, in a few experimental modules, breeding tritium. Nb 3 Sn and NbTi superconductors are used in the toroidal field coil design. The coil system was developed for a maximum field of 11 T at the winding (to give a field on axis of 4.8 T), and combines multifilamentary superconducting cable with forced flow of supercritical helium enclosed in a steel conduit. The structural system uses a stainless steel center bucking ring and intercoil box beam bracing to provide rigid support for coils against the centering force, overturning moments from poloidal fields and faults, other external forces, and thermal stresses. The poloidal magnetics system is specially designed both to reduce the total volt-second energy requirements and to reduce the magnitude of the rate of field change at the toroidal field coils. The rate of field change imposed upon the toroidal field coils is reduced by at least a factor of 3.3 compared to that due to the plasma alone. Tritium processing, tritium containment and vacuum systems employ double containment and atmospheric cleanup to minimize releases. The document also contains discussions of systems integration and assembly, key research and development needs, and schedule considerations

  5. Cryogenic structures of superconducting coils for fusion experimental reactor 'ITER'

    International Nuclear Information System (INIS)

    Nakajima, Hideo; Iguchi, Masahide; Hamada, Kazuya; Okuno, Kiyoshi; Takahashi, Yoshikazu; Shimamoto, Susumu

    2013-01-01

    This paper describes both structural materials and structural design of the Toroidal Field (TF) coil and Central Solenoid (CS) for the International Thermonuclear Experimental Reactor (ITER). All the structural materials used in the superconducting coil system of the ITER are austenitic stainless steels. Although 316LN is used in the most parts of the superconducting coil system, the cryogenic stainless steels, JJ1 and JK2LB, which were newly developed by the Japan Atomic Energy Agency (JAEA) and Japanese steel companies, are used in the highest stress area of the TF coil case and the whole CS conductor jackets, respectively. These two materials became commercially available based on demonstration of productivity and weldability of materials, and evaluations of 4 K mechanical properties of trial products including welded parts. Structural materials are classified into five grades depending on stress distribution in the TF coil case. JAEA made an industrial specification for mass production based on the ITER requirements. In order to simplify quality control in mass production, JAEA has used materials specified in the material section of 'Codes for Fusion Facilities - Rules on Superconducting Magnet Structure (2008)' issued by the Japan Society of Mechanical Engineers (JSME) in October 2008, which was established using an extrapolation method of 4 K material strengths from room temperature strength and chemical compositions developed by JAEA. It enables steel suppliers to easily control the quality of products at room temperature. JAEA has already started actual production with several manufacturing companies. The first JJ1 product to be used in the TF coil case and the first JK2LB jackets for CS were completed in October and September 2013, respectively. (author)

  6. Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

  7. The Japan Power Demonstration Reactor dismantling project. Radiation control

    International Nuclear Information System (INIS)

    Tomii, Hiroyuki; Seiki, Yoshihiro

    1996-01-01

    In the Japan Power Demonstration Reactor (JPDR) dismantling project, radiation control was performed properly with routine and special monitoring to keep the occupational safety and to collect data necessary for future dismantling of nuclear facilities. This report describes a summary of radiation control in the dismantling activities and some results of parametric analysis on dose equivalent evaluation, and introduces the following knowledge on radiological protection effectiveness of the dismantling systems applied in the project. a) Use of remote dismantling systems was effective in reducing equivalent workplace exposure. b) Utilization of existing facilities as radiation shield or radioactivity containment was effective in reducing workplace exposure, and also in increasing work efficiency. c) Use of underwater cutting systems was useful to minimize air contamination, and to reduce the dose equivalent rate in the working area. d) In the planning of dismantling, it is necessary to optimize the radiation protection by analyzing dismantling work procedures and evaluating radiological features of the dismantling systems applied, including additional work which the systems require brought from such activities. (author)

  8. 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.

  9. 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)

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

    2008-01-01

    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

  10. Current status and perspective of advanced loop type fast reactor in fast reactor cycle technology development project

    International Nuclear Information System (INIS)

    Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

    2007-01-01

    After selecting the combination of the sodium-cooled fast reactor (SFR) with oxide fuel, the advanced aqueous reprocessing and the simplified pelletizing fuel fabrication as the most promising concept of FR cycle system, 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was finalized in 2006. Instead, a new project, Fast Reactor Cycle Technology Development Project (FaCT Project) was launched in Japan focusing on development of the selected concepts. This paper describes the current status and perspective of the advanced loop type SFR system in the FaCT Project, especially on the design requirements, current design as well as the related innovative technologies together with the development road-map. Some considerations on advantages of the advanced loop type design are also described. (authors)

  11. International project on innovative nuclear reactors and fuel cycles (INPRO)

    International Nuclear Information System (INIS)

    Omoto, A.

    2006-01-01

    The IAEA's project INPRO was initiated in order to provide a forum for discussion of experts and policy makers on all aspects of nuclear energy planning as well as on the development and deployment of innovative nuclear energy systems (INS). It brings together technology holders users and potential users to consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles, but it pays particular attention to the needs of developing countries. Currently INPRO members count 24 including even three countries, which are not yet operating nuclear reactors. Its initial phase has produced an outlook into the future of the energy markets and defined basic principles, user requirements and criteria in the following areas as TECDOC1362 in June 2003; Economics, Environment, Fuel Cycle and Waste, Safety, Proliferation Resistance and Crosscutting Issues. This assessment methodology can be applied for screening an INS, comparing different INS to find a preferred INS consistent with the needs of a given state, and identifying RD and D needs. The methodology has be validated through case studies and updated as TECDOC1434 in December 2004. Currently, besides producing a manual for each chapter of TECDOC1434, six assessment studies of various INS options are being carried out and the number of such studies is increasing. Further several tasks are ongoing including modeling and analysis of global and regional balance of resources and INS deployment scenarios in order to gain the better perspective of future implication of INS deployment as well as to identify challenges and opportunities of INS. It is envisioned that INPRO will continue to develop with three planned major pillars of activity; methodology, infrastructure and coordination for planning of R and D activities. The paper discusses the progress and status of INPRO as well as the future prospect of INPRO activities

  12. Experimental and numerical investigation of bubble column reactors

    NARCIS (Netherlands)

    Bai, W.

    2010-01-01

    Due to various advantages, such as simple geometry, ease of operation, low operating and maintenance costs, excellent heat and mass transfer characteristics, bubble column reactors are frequently used in chemical, petrochemical, biochemical, pharmaceutical, metallurgical industries for a variety of

  13. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Energy Technology Data Exchange (ETDEWEB)

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  14. Experimental study of the passive flooding system in the WWER-1000 reactor

    International Nuclear Information System (INIS)

    Malyshev, A.B.; Efanov, A.D.; Kalyakin, S.G.

    2002-01-01

    The design solution of the passive flooding system in the WWER-1000 reactor core with the V-392 reactor facility and the scheme of the GE-2 large-scale thermohydraulic stand for substantiation of its functions are presented. The proposals, improving the efficiency of the system are developed on the basis of the experimental studies on the equipment input-output operational characteristics and the recommendations on the substantiation of the function of the reactor core flooding system are given [ru

  15. Review of accident analyses of RB experimental reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2003-01-01

    The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA 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 U0 2 , 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 VINCA Institute, an independent regulator)' 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. (author)

  16. 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.

  17. The Puerto Rico nuclear center reactor conversion project

    Energy Technology Data Exchange (ETDEWEB)

    Brown-Campos, R [Puerto Rico Nuclear Center (Puerto Rico)

    1974-07-01

    For the purpose of upgrading the control and instrumentation system to meet new AEC requirements, to increase the available neutron flux for experimenters and to replace burned out fuel the Puerto Rico Nuclear Center started a modification program on its old MTR type, one megawatt reactor on March 1971. A TRIGA core utilizing the newly developed FLIP fuel, capable of operating at two megawatts with natural convection cooling and with pulsing capabilities was chosen. The major conversion tasks included: 1. Modification of the bridge, tower and grid plate structures, 2. Modification of the water cooling system (inside the reactor pool), 3. Installation of a larger heat exchanger and cooling tower, 4. Installation of a new instrumentation and control console (including neutron detectors and rod drive mechanisms). 5. Installation of a TRIGA FLIP core. Initial criticality was achieved on January 1972. For the chosen operating configuration the critical mass was 11,522 grams of uranium 235. Core excess reactivity was $7.12 and the total (5) rod worth was $12.06. During the early stages of the startup program to determine the basic core parameters and while conducting a stepwise increase in power to the design power level of two megawatts a power fluctuation on all neutron detectors was noticed. It was determined that the power fluctuations started at about 1.4 megawatts and sharply increased as power approached 2 megawatts. Experiments to determine the cause of the problem and to correct the condition were conducted on July and December 1972 and June 1973. Modifications to the core included changing fuel pin pitch and the addition of dummy elements in the central region of the core. Final acceptance by AEC Headquarters was requested on October 1973. (author)

  18. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  19. Design of an Experimental Facility for Passive Heat Removal in Advanced Nuclear Reactors

    Science.gov (United States)

    Bersano, Andrea

    With reference to innovative heat exchangers to be used in passive safety system of Gen- eration IV nuclear reactors and Small Modular Reactors it is necessary to study the natural circulation and the efficiency of heat removal systems. Especially in safety systems, as the decay heat removal system of many reactors, it is increasing the use of passive components in order to improve their availability and reliability during possible accidental scenarios, reducing the need of human intervention. Many of these systems are based on natural circulation, so they require an intense analysis due to the possible instability of the related phenomena. The aim of this thesis work is to build a scaled facility which can reproduce, in a simplified way, the decay heat removal system (DHR2) of the lead-cooled fast reactor ALFRED and, in particular, the bayonet heat exchanger, which transfers heat from lead to water. Given the thermal power to be removed, the natural circulation flow rate and the pressure drops will be studied both experimentally and numerically using the code RELAP5 3D. The first phase of preliminary analysis and project includes: the calculations to design the heat source and heat sink, the choice of materials and components and CAD drawings of the facility. After that, the numerical study is performed using the thermal-hydraulic code RELAP5 3D in order to simulate the behavior of the system. The purpose is to run pretest simulations of the facility to optimize the dimensioning setting the operative parameters (temperature, pressure, etc.) and to chose the most adequate measurement devices. The model of the system is continually developed to better simulate the system studied. High attention is dedicated to the control logic of the system to obtain acceptable results. The initial experimental tests phase consists in cold zero power tests of the facility in order to characterize and to calibrate the pressure drops. In future works the experimental results will be

  20. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.-P.; Ritter, S

    2009-03-15

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  1. Environmentally-assisted cracking in austenitic light water reactor structural materials. Final report of the KORA-I project

    International Nuclear Information System (INIS)

    Seifert, H.-P.; Ritter, S.

    2009-03-01

    The following document is the final report of the KORA-I project, which was performed at the Paul Scherrer Institute (PSI) between 2006 and 2008 and was funded by the Swiss Nuclear Safety Inspectorate (ENSI). The three sub-projects of KORA-I covered the experimental characterisation of the effect of the reactor coolant environment on fatigue initiation and crack growth in austenitic stainless steels under boiling and pressurised water reactor conditions, the experimental evaluation of the potential and limits of the electrochemical noise measurement technique for the early detection of stress corrosion cracking initiation in austenitic stainless steels under boiling water reactor/normal water chemistry conditions, as well as the characterisation of the stress corrosion crack growth behaviour in the fusion line region of an Alloy 182-low-alloy reactor pressure vessel steel dissimilar metal weld. The main scientific results and major conclusions of the three sub-projects are discussed in three independent parts of this report. (author)

  2. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

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

    NARCIS (Netherlands)

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

    2005-01-01

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

  4. A dense Pd/Ag membrane reactor for methanol steam reforming: Experimental study

    NARCIS (Netherlands)

    Basile, A.; Gallucci, F.; Paturzo, L.

    2005-01-01

    This paper focuses on an experimental study of the methanol steam reforming (MSR) reaction. A dense Pd/Ag membrane reactor (MR) has been used, and its behaviour has been compared to the performance of a traditional reactor (TR) packed with the same catalyst type and amount. The parameters

  5. Experimental Methods Related to Coupled Fast-Thermal Systems at the RB Reactor

    International Nuclear Information System (INIS)

    Pesic, M.

    2002-01-01

    In addition to the review of RB reactor characteristics this presentation is focused on the coupled fast-thermal systems achieved at the reactor. The following experimental methods are presented: neutron spectra measurements; steady state experiments and kinetic measurements ( β eff ) related to the coupled fast-thermal cores

  6. 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

  7. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  8. Thermochemical storage for CSP via redox structured reactors/heat exchangers: The RESTRUCTURE project

    Science.gov (United States)

    Karagiannakis, George; Pagkoura, Chrysoula; Konstandopoulos, Athanasios G.; Tescari, Stefania; Singh, Abhishek; Roeb, Martin; Lange, Matthias; Marcher, Johnny; Jové, Aleix; Prieto, Cristina; Rattenbury, Michael; Chasiotis, Andreas

    2017-06-01

    The present work provides an overview of activities performed in the framework of the EU-funded collaborative project RESTRUCTURE, the main goal of which was to develop and validate a compact structured reactor/heat exchanger for thermochemical storage driven by 2-step high temperature redox metal oxide cycles. The starting point of development path included redox materials qualification via both theoretical and lab-scale experimental studies. Most favorable compositions were cobalt oxide/alumina composites. Preparation of small-scale structured bodies included various approaches, ranging from perforated pellets to more sophisticated honeycomb geometries, fabricated by extrusion and coating. Proof-of-concept of the proposed novel reactor/heat exchanger was successfully validated in small-scale structures and the next step included scaling up of redox honeycombs production. Significant challenges were identified for the case of extruded full-size bodies and the final qualified approach related to preparation of cordierite substrates coated with cobalt oxide. The successful experimental evaluation of the pilot reactor/heat exchanger system constructed motivated the preliminary techno-economic evaluation of the proposed novel thermochemical energy storage concept. Taking into account experimental results, available technologies and standard design aspects a model for a 70.5 MWe CSP plant was defined. Estimated LCOE costs were calculated to be in the range of reference values for Combined Cycle Power Plants operated by natural gas. One of main cost contributors was the storage system itself, partially due to relatively high cost of cobalt oxide. This highlighted the need to identify less costly and equally efficient to cobalt oxide redox materials.

  9. Decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East. Project final report

    International Nuclear Information System (INIS)

    Fellhauer, C.R.; Clark, F.R.

    1997-10-01

    The decontamination and dismantlement of the JANUS Reactor at Argonne National Laboratory-East (ANL-E) was completed in October 1997. Descriptions and evaluations of the activities performed and analyses of the results obtained during the JANUS D and D Project are provided in this Final Report. The following information is included: objective of the JANUS D and D Project; history of the JANUS Reactor facility; description of the ANL-E site and the JANUS Reactor facility; overview of the D and D activities performed; description of the project planning and engineering; description of the D and D operations; summary of the final status of the JANUS Reactor facility based upon the final survey results; description of the health and safety aspects of the project, including personnel exposure and OSHA reporting; summary of the waste minimization techniques utilized and total waste generated by the project; and summary of the final cost and schedule for the JANUS D and D Project

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

    KAUST Repository

    Wang, Jui-Yang

    2017-01-01

    levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental

  11. Chemical looping reforming in packed-bed reactors : modelling, experimental validation and large-scale reactor design

    NARCIS (Netherlands)

    Spallina, V.; Marinello, B.; Gallucci, F.; Romano, M.C.; van Sint Annaland, M.

    This paper addresses the experimental demonstration and model validation of chemical looping reforming in dynamically operated packed-bed reactors for the production of H2 or CH3OH with integrated CO2 capture. This process is a combination of auto-thermal and steam methane reforming and is carried

  12. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  13. Action Memorandum for Decommissioning the Engineering Test Reactor Complex under the Idaho Cleanup Project

    International Nuclear Information System (INIS)

    A. B. Culp

    2007-01-01

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared and released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessel. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface

  14. Small and medium power reactors: project initiation study, Phase 1

    International Nuclear Information System (INIS)

    1985-07-01

    In conformity with the Agency's promotional role in the peaceful uses of nuclear energy, IAEA has provided, over the past 20 years, assistance to Member States, particularly developing countries, in planning for the introduction of nuclear power plants in the Small and Medium range (SMPR). However these efforts did not produce any significant results in the market introduction of these reactors, due to various factors. In 1983 the Agency launched a new SMPR Project Initiation Study with the objective of surveying the available designs, examining the major factors influencing the decision-making processes in Developing Countries and thereby arriving at an estimate of the potential market. Two questionnaires were used to obtain information from possible suppliers and prospective buyers. The Nuclear Energy Agency of OECD assisted in making a study of the potential market in industrialized countries. The information gained during the study and discussed during a Technical Committee Meeting on SMPRs held in Vienna in March 1985, along with the contribution by OECD-NEA is embodied in the present report

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

    International Nuclear Information System (INIS)

    1985-12-01

    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)

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

    International Nuclear Information System (INIS)

    Miki, Nobuharu; Iida, Fumio; Wachi, Yoshihiro; Toyoda, Katsuyoshi; Hashizume, Takashi; Konno, Masayuki.

    1988-06-01

    This report describes the FER magnet design which was conducted last year (1987). Based on a large uncertainty of the physics assumption, two sets of FER concepts have been developed. One is based on the best existing physics data bases and another is based on rather conservative physics bases. In the magnet design, the improvements of superconducting magnet design were investigated to reduce the reactor size and to realize higher reactor-core performance. In addition, we studied several critical technical issues that affect the magnet design specification. (author)

  17. Using harmonical analysis for experimental verification of reactor dynamics

    International Nuclear Information System (INIS)

    Hrstka, V.

    1974-01-01

    The questions are discussed of the accuracy of the method of static programming when applied to digital harmonic analysis, with regard to the variation of the mean value of the analyzed signals, and to the use of symmetrical trapezoidal periodical signals. The evaluation is made of the suitability of the above-mentioned method in determining the frequency characteristic of the SR-OA reactor. The results obtained were applied to planning the start-up experiments of the KS-150 reactor at the A-1 nuclear power station. (author)

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

    International Nuclear Information System (INIS)

    Kobayashi, Takeshi; Yamada, Masao; Mizoguchi, Tadanori

    1987-09-01

    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)

  19. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

  20. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

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

    International Nuclear Information System (INIS)

    Petti, D.A.; Haire, J.C.

    1993-12-01

    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

  2. The Experimental Role of Accounting in Shaping Project

    DEFF Research Database (Denmark)

    Christensen, Mark; Skærbæk, Peter; Tryggestad, Kjell

    2016-01-01

    , 2009) and the literature on accounting as programs (Miller, 1991, Miller and Power, 2013) by showing how accounting plays an experimental role in shaping the dynamics between a New Public Management programme of outsourcing and its projects. The study illuminates decisionmaking about programs...... and strategic options as a trial of strength involving accounting in an experimental role, in our case, over how to source the facilities management in the Danish Defence Forces. Our contribution is to show how accounting plays an experimental role in generating new and unexpected program options...... of economic experiments with (out-)sourcing options using various accounting data such as; location specific cost data in vivo (real life data like those from the accounting system) for individual barracks, and aggregated cost data and cost projections in vitro ( in a controlled experimental project setting...

  3. General description of preliminary design of an experimental fusion reactor and the future problems

    International Nuclear Information System (INIS)

    Sako, Kiyoshi

    1976-01-01

    Recently, the studies on plasma physics has progressed rapidly, and promising experimental data emerged successively. Especially expectation mounts high that Tokamak will develop into power reactors. In Japan, the construction of large plasma devices such as JT-60 of JAERI is going to start, and after several years, the studies on plasma physics will come to the end of first stage, then the main research and development will be directed to power reactors. The studies on the design of practical fusion reactors have been in progress since 1973 in JAERI, and the preliminary design is being carried out. The purposes of the preliminary design are the clarification of the concept of the experimental reactor and the requirements for the studies on core plasma, the examination of the problems for developing main components and systems of the reactor, and the development of design technology. The experimental reactor is the quasi-steady reactor of 100 MW fusion reaction output, and the conditions set for the design and the basis of their setting are explained. The outline of the design, namely core plasma, blankets, superconductive magnets and the shielding with them, vacuum wall, neutral particle injection heating device, core fuel supply and exhaust system, and others, is described. In case of scale-up the reactor structural material which can withstand neutron damage must be developed. (Kako, I.)

  4. Aqueous homogeneous suspension reactor project. Report over the year 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

    During 1975, reactor power has been increased to the design power of 1000 KW, whereas power fluctuations show a decrease with increased mean power. Operation experience with the reactor and associated instrumentation during 1975 is described. The results of the experiments done for fuel irradiations and investigations in the KSTR fuel, mainly to determine the amount of erosion products on the fuel, are described. Concerning the safety of operation of the KSTR reactor, several actions had to be taken mainly to replace or repair components or instrumentation of the reactor. Radiological safety and radioactivity discharges during 1975 are reported.

  5. The IAEA International Project on Innovative Reactors and Fuel Systems

    International Nuclear Information System (INIS)

    Mourogov, V.M.

    2001-01-01

    and policy makers. A 2000 the IAEA General Conference resolution invited 'all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology'. In response to this invitation, the IAEA initiated an 'International Project on Innovative Nuclear Reactors and Fuel Cycles', INPRO. The INPRO Project will be implemented in two phases. In the first phase, the main objective is to identify user requirements facilitating large scale nuclear energy development in the 21st century in the following areas: Resources, Demand and Economics; Environment, Spent Fuel and Waste; Safety, and Non-proliferation. Plus two crosscutting groups addressing Criteria and Methodology; and Institutional, Infrastructure, Social and Sustainability Requirements. Upon successful completion of the first phase, taking into account advice from the Steering Committee, and with the approval of participating Member States, a second phase of INPRO may be initiated. It would examine, in the context of available technologies, the feasibility of an international project including the identification of technologies that might appropriately be implemented by Member States within such an international project. We believe that INPRO'S global character, encompassing both designers and end users and their user's requirements, its long time horizon, its consideration of the changing energy sector and its broad based input through IAEA membership all will make it a valuable forum for the assessment of perspectives for nuclear in the 21st century. (author)

  6. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  7. Simulation test of PIUS-type reactor with large scale experimental apparatus

    International Nuclear Information System (INIS)

    Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

    1995-01-01

    A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

  8. Some particular problems put by operating experimental reactors

    International Nuclear Information System (INIS)

    Candiotti, C.; Mabeix, R.; Uguen, R.

    1960-01-01

    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) [fr

  9. Linear and nonlinear stability analysis, associated to experimental fast reactors

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    Phenomena associated to the physics of fast neutrons were analysed by linear and nonlinear Kinetics with arbitrary feedback. The theoretical foundations of linear kinetics and transfer functions aiming at the analysis of fast reactors stability, are established. These stability conditions were analitically proposed and investigated by digital and analogic programs. (E.G.) [pt

  10. Engineering safety features for high power experimental reactors

    International Nuclear Information System (INIS)

    Doval, A.; Villarino, E.; Vertullo, A.

    2000-01-01

    In the present analysis we will focus our attention in the way engineering safety features are designed in order to prevent fuel damage in case of abnormal or accidental situations. To prevent fuel damage two main facts must be considered, the shutdown of the reactor and the adequate core cooling capacity, it means that both, neutronic and thermohydraulic aspects must be analysed. Some neutronic safety features are common to all power ranges like negative feedback reactivity coefficients and the required number of control rods containing the proper absorber material to shutdown the reactor. From the thermohydraulic point of view common features are siphon-breaker devices and flap valves for those powers requiring cooling in the forced convection regime. For the high power reactor group, the engineering safety features specially designed for a generic reactor of 20 MW, will be presented here. From the neutronic point of view besides the common features, and to comply with our National Regulatory Authority, a Second Shutdown System was designed as a redundant shutdown system in case the control plates fail. Concerning thermohydraulic aspects besides the pump flywheels and the flap valves providing the natural convection loop, a metallic Chimney and a Chimney Water Injection System were supplied. (author)

  11. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    Hohara, Shinya

    2015-01-01

    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.)

  12. IAEA coordinated research project (CRP) on 'Analytical and experimental benchmark analyses of accelerator driven systems'

    International Nuclear Information System (INIS)

    Abanades, Alberto; Aliberti, Gerardo; Gohar, Yousry; Talamo, Alberto; Bornos, Victor; Kiyavitskaya, Anna; Carta, Mario; Janczyszyn, Jerzy; Maiorino, Jose; Pyeon, Cheolho; Stanculescu, Alexander; Titarenko, Yury; Westmeier, Wolfram

    2008-01-01

    In December 2005, the International Atomic Energy Agency (IAEA) has started a Coordinated Research Project (CRP) on 'Analytical and Experimental Benchmark Analyses of Accelerator Driven Systems'. The overall objective of the CRP, performed within the framework of the Technical Working Group on Fast Reactors (TWGFR) of IAEA's Nuclear Energy Department, is to increase the capability of interested Member States in developing and applying advanced reactor technologies in the area of long-lived radioactive waste utilization and transmutation. The specific objective of the CRP is to improve the present understanding of the coupling of an external neutron source (e.g. spallation source) with a multiplicative sub-critical core. The participants are performing computational and experimental benchmark analyses using integrated calculation schemes and simulation methods. The CRP aims at integrating some of the planned experimental demonstration projects of the coupling between a sub-critical core and an external neutron source (e.g. YALINA Booster in Belarus, and Kyoto University's Critical Assembly (KUCA)). The objective of these experimental programs is to validate computational methods, obtain high energy nuclear data, characterize the performance of sub-critical assemblies driven by external sources, and to develop and improve techniques for sub-criticality monitoring. The paper summarizes preliminary results obtained to-date for some of the CRP benchmarks. (authors)

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

    International Nuclear Information System (INIS)

    Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

    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

  14. 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.

  15. Experimental Boiling Water Reactor decontamination and decommissioning project

    International Nuclear Information System (INIS)

    Fellhauer, C.

    1995-01-01

    The author begins by discussing the problems encountered during decontamination and decommissioning. Next, he discusses waste packaging and recycling. His last topic of lessons learned is subdivided into prevention and early detection, recovery issues, management issues, and noteworthy practices

  16. The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

    International Nuclear Information System (INIS)

    Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

    1977-01-01

    The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

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

    International Nuclear Information System (INIS)

    1984-12-01

    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 950 0 C. 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)

  18. Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

    The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

  19. Staffing requirements for future small and medium reactors based on projections in the Russian Federation

    International Nuclear Information System (INIS)

    Antonovsky, G.M.; Kodochigov, N.G.; Kurachenkov, A.V.; Novikov, V.V.

    2001-01-01

    Experimental Design Bureau of Mechanical Engineering (OKBM) specializes in the development of small and medium power reactors having different purposes. They include reactor plants for NPHPP, nuclear district heating power plants and propulsion plants. Small and medium power plants have simpler processes of electricity and heat production, less systems, simpler control algorithms and considerably enhanced inherent safety properties. These plants are mainly equipped with passive safety systems. These properties are especially characteristic for reactor plants of nuclear district heating power plants and HTG reactor plants. The designs of small and medium power plants actually provide a high degree of control automation which considerably reduces workload on the personnel in both normal and abnormal operation conditions. All this allows the reduction in personnel for small and medium power reactors if compared to high capacity reactor plants. But due to objective reasons the specific number of personnel (man/MW) for average and especially small capacity reactors considerably exceeds the value for high capacity reactor plants. At the same time one can propose a set of organization - technical measures allowing the increase in this value in future. Safety requirements imposed for small and average capacity reactors are the same or more strict than those for high capacity reactors. That's why the requirements to the training of personnel for such reactor plants are not allowed to be lowered if compared to the requirements imposed to the personnel of high capacity reactors. (author)

  20. Job task and functional analysis of the Division of Reactor Projects, office of Nuclear Reactor Regulation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Morzinski, J.A.; Gilmore, W.; Hahn, H.A.

    1998-07-10

    A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.

  1. EXPERIMENTAL EVALUATION OF THE FULLY LOADED ELK RIVER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, J. R.; Diaz, A.

    1963-06-15

    The loading and testing program of the Elk River Reactor confirmed the predicted values. The measured cold, clean excess reactivity agrees to 2% and the control rod worths to 1% of the calculated values. The reactivity for various core loadings and rod positions is tabulated. The effects of spiked elements on the reactivity and radial peak-toaverage power ratio were studied. (D.L.C.)

  2. Storage of plugs and experimental devices from reactors

    International Nuclear Information System (INIS)

    Cerre, P.; Mestre, E.

    1961-01-01

    - Within the general programme of storage and treatment of radioactive waste produced by the various operations carried out in an atomic center, it is useful to consider separately the problem of certain waste from reactors, which, because of its size and physical nature, has to be stored with a view to being later treated and finally evacuated. The solution which we propose for this storage problem is presented in this paper. (authors) [fr

  3. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  4. Design of project management system for 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Zhu Yan; Xu Yuanhui

    1998-01-01

    A framework of project management information system (MIS) for 10 MW high temperature gas-cooled test reactor is introduced. Based on it, the design of nuclear project management information system and project monitoring system (PMS) are given. Additionally, a new method of developing MIS and Decision Support System (DSS) has been tried

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

    Science.gov (United States)

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

    2017-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kruijf, W.J.M. de E-mail: kruijf@iri.tudelft.nl; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C

    2003-04-01

    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.

  7. 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)

    Kruijf, W.J.M. de; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C.

    2003-01-01

    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

  8. Plan for Demonstration of Online Monitoring for the Light Water Reactor Sustainability Online Monitoring Project

    Energy Technology Data Exchange (ETDEWEB)

    Magdy S. Tawfik; Vivek Agarwal; Nancy J. Lybeck

    2011-09-01

    Condition based online monitoring technologies and development of diagnostic and prognostic methodologies have drawn tremendous interest in the nuclear industry. It has become important to identify and resolve problems with structures, systems, and components (SSCs) to ensure plant safety, efficiency, and immunity to accidents in the aging fleet of reactors. The Machine Condition Monitoring (MCM) test bed at INL will be used to demonstrate the effectiveness to advancement in online monitoring, sensors, diagnostic and prognostic technologies on a pilot-scale plant that mimics the hydraulics of a nuclear plant. As part of this research project, INL will research available prognostics architectures and their suitability for deployment in a nuclear power plant. In addition, INL will provide recommendation to improve the existing diagnostic and prognostic architectures based on the experimental analysis performed on the MCM test bed.

  9. Experimental results on advanced inertial fusion schemes obtained within the HiPER project

    International Nuclear Information System (INIS)

    Batani, Dimitri; Santos, Jorge J.; Schurtz, Guy; Hulin, Sebastien; Ribeyre, Xavier; Nicolai, Philippe; Vauzour, Benjamin; Dorchies, Fabien; Gizzi, Leonida A.; Koester, Petra; Labate, Luca; Honrubia, Javier; Antonelli, Luca; Morace, Alessio; Volpe, Luca; Nazarov, Wiger; Pasley, John; Richetta, Maria; Lancaster, Kate; Spindloe, Christopher; Tolley, Martin; Neely, David; Kozlova, Michaela; Nejdl, Jaroslav; Rus, Bedrich; Wolowski, Jerzy; Badziak, Jan

    2012-01-01

    This paper presents the results of experiments conducted within the Work Package 10 (fusion experimental programme) of the HiPER project. The aim of these experiments was to study the physics relevant for advanced ignition schemes for inertial confinement fusion, i.e. the fast ignition and the shock ignition. Such schemes allow to achieve a higher fusion gain compared to the indirect drive approach adopted in the National Ignition Facility in United States, which is important for the future inertial fusion energy reactors and for realising the inertial fusion with smaller facilities. (authors)

  10. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

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

  11. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

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

    International Nuclear Information System (INIS)

    2012-08-01

    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.

  13. 105-H Reactor Interim Safe Storage Project Final Report

    International Nuclear Information System (INIS)

    Ison, E.G.

    2008-01-01

    The following information documents the decontamination and decommissioning of the 105-H Reactor facility, and placement of the reactor core into interim safe storage. The D and D of the facility included characterization, engineering, removal of hazardous and radiologically contaminated materials, equipment removal, decontamination, demolition of the structure, and restoration of the site. The ISS work also included construction of the safe storage enclosure, which required the installation of a new roofing system, power and lighting, a remote monitoring system, and ventilation components.

  14. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    International Nuclear Information System (INIS)

    Cagle, C.D.

    1982-10-01

    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

  15. Experimental researches of nuclear reactor neutron and gamma radiation scattering into the atmosphere

    International Nuclear Information System (INIS)

    Istomin, Yu.L.; Zelensky, D.I.; Cherepnin, Yu.S.; Orlov, Yu.V.; Netecha, M.E.; Avaev, V.N.; Vasel'ev, G.A.; Sakamoto, H.; Nomura, Y.; Naito, Y.

    1998-01-01

    In the report there are results of measuring radiation distribution on the caps of the RA and IWG.1M research reactors. Comparative analysis of the results is also in the report. There are neutron spectra in the interval of energies from 10 -9 to 13 MeV above RA and IWG.1M reactors. The spectra were measured with a set of activation detectors. Measurements were calculated to a nominal rate: for RA reactor - 300 kw, for IWG.1M - 7 MW. Thus, in the course of the experiment, vast experimental information relating to distribution of the RA and IWG.1M reactor gamma and neutron radiation scattered in the air for distances varying from 50 to 1000 m from the reactors has become available. The data obtained are to be used to verify the calculation codes and to validate the group nuclear constants

  16. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  17. EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

    International Nuclear Information System (INIS)

    Mott, J.E.; Radtke, W.H.; King, R.W.

    1986-02-01

    The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

  18. Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

    International Nuclear Information System (INIS)

    Rana, Y.S.; Mishra, Abhishek; Singh, Tej

    2016-06-01

    Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

  19. Experimental Irradiations of Materials and Fuels in the BR2 Reactor: An Overview of Current Programmes

    International Nuclear Information System (INIS)

    Van Dyck, S.; Koonen, E.; Verwerft, M.; Wéber, M.

    2013-01-01

    The BR2 material test reactor offers a variety of experimental irradiation possibilities for testing of materials, fuels and instruments. The current paper gives an overview of the recent and ongoing programmes in order to illustrate the experimental potential of the reactor. Three domains of applications are reviewed: Irradiation of materials and fuels for pressurised water reactors (PWR); irradiation of materials for accelerator driven systems (ADS), cooled by liquid lead alloys; and irradiation of fuel for Material Test Reactors (MTR). For PWR relevant tests, a dedicated loop is available, providing a full simulation of the thermo hydraulic conditions of a PWR. ADS related tests require particular control of the irradiation environment and the necessary safety precautions in order to avoid 210 Po contamination. In-core mechanical testing of materials is done in comparison and complimentarily to post-irradiation examinations in order to assess flux related effects on the deformation behaviour of materials. (author)

  20. Completion of the experimental equipment systems and preparation of practical tutorials on the Dalat Nuclear Research Reactor for nuclear science and technology education

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Luong Ba Vien; Nguyen Minh Tuan; Nguyen Kien Cuong; Pham Quang Huy; Tran Tri Vien

    2015-01-01

    The project Completion of the experimental equipment systems and preparation of practical tutorials on the Dalat Nuclear Research Reactor for nuclear science and technology education performed by Dalat Nuclear Research Institute and financed by Ministry of Science and Technology aimed at strengthening the training capability of nuclear human resources. The content of this work includes: i) Improvement of experimental equipment; ii) Compilation of training material for experiments with the improved equipment systems on the reactor; iii) Compilation of training material for reactor calculations includes the following areas: neutronics, hydrothermal, safety analysis and accident consequence analysis. Results of the project provide important conditions to support practical educational and training curriculums in nuclear science and technology. (author)

  1. Hefei experimental hybrid fusion-fission reactor conceptual design

    International Nuclear Information System (INIS)

    Qiu Lijian; Luan Guishi; Xu Qiang

    1992-03-01

    A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

  2. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  3. The activities of the committee 'Kernreaktorregelung' (nuclear reactor control) in the past few years and future projects

    International Nuclear Information System (INIS)

    Knecht, O.

    1976-01-01

    Results achieved so far and future projects are portrayed in detail: 1) VDI/VDE 3527-Graphical symbols for nuclear reactor control; 2) VDI/VDE 3528-Special terms and definitons for nuclear reactor control; 3) 8 data sheets on reactor control; 4) VDI/VDE 3530-Characterisation of reactor control rod drives. (orig./HP) [de

  4. Fuel requirements for experimental devices in MTR reactors. A perturbation model for reactor core analysis

    International Nuclear Information System (INIS)

    Beeckmans de West-Meerbeeck, A.

    1991-01-01

    Irradiation in neutron absorbing devices, requiring high fast neutron fluxes in the core or high thermal fluxes in the reflector and flux traps, lead to higher density fuel and larger core dimensions. A perturbation model of the reactor core helps to estimate the fuel requirements. (orig.)

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

    International Nuclear Information System (INIS)

    Amorim, E.S. do; Moura Neto, C. de; Rosa, M.A.P.

    1980-07-01

    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.) [pt

  6. Analytical prediction and experimental verification of reactor safety system injection transient

    International Nuclear Information System (INIS)

    Roy, B.N.; Nomm, E.

    1991-01-01

    This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs

  7. Scram reliability under seismic conditions at the Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Roglans, J.; Wang, C.Y.; Hill, D.J.

    1993-01-01

    A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

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

    International Nuclear Information System (INIS)

    Calamand, D.; Wouters, R. de; Knipe, A.D.; Menil, R.

    1984-10-01

    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

  9. Progress and status of the international project on innovative nuclear reactors and fuel cycles (INPRO) - 5182

    International Nuclear Information System (INIS)

    Ponomarev, A.; Fesenko, G.; Grigoriev, F.G.; Korinny, A.; Phillips, J.R.; Rho, K.

    2015-01-01

    The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 through IAEA General Conference resolution. INPRO cooperates with Member States to ensure that sustainable nuclear energy is available to help meet the energy needs of the 21. century. INPRO membership has grown to 41 members and 16 observers. The paper presents the current prospectus of the INPRO programme and details the most recent achievements in the following 7 projects: 1) the GAINS project (Global Architecture of Innovative Nuclear Energy Systems with thermal and fast reactors and a closed nuclear fuel cycle); 2) the SYNERGIES project applies and amends the analytical framework developed in GAINS project to examine more specifically the various forms of regional collaboration among nuclear energy suppliers and users; 3) the KIND project (Key Indicators for Innovative Nuclear Energy Systems) has the objective of developing guidance on the evaluation on innovative nuclear technologies; 4) the ROADMAPS project addresses several possible stages toward nuclear energy sustainability; 5) the RISC project aims at demonstrating that the evolution of safety requirements and technical innovations provide continual progress towards the avoidance of evacuation measures outside NPP sites in case of severe accidents; 6) the FANES project has the objective of carrying out feasibility analyses of advanced and innovative fuels for different reactor systems; and 7) the WIRAF project aims at identifying problematic waste from innovative reactor designs and corresponding nuclear fuel cycles

  10. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  11. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  12. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bosland, L. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France)], E-mail: loic.bosland@irsn.fr; Funke, F. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany); Girault, N. [Institut de Radioprotection et de Surete Nucleaire, DPAM, SEMIC, LETR - CEN Cadarache, BP 3, 13115 Saint Paul Lez Durance (France); Langrock, G. [AREVA NP GmbH, Technical Center, P.O. Box 1109, D-91001 Erlangen (Germany)

    2008-12-15

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two

  13. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Bosland, L.; Funke, F.; Girault, N.; Langrock, G.

    2008-01-01

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two different kinetic models

  14. ARCHER Project: Progress on Material and component activities for the Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.

    2014-01-01

    The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) integrated project is a four year project which was started in 2011 as part of the European Commission 7th Framework Programme (FP7) to perform High Temperature Reactor technology R&D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research & Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on ARCHER materials and component activities since the start of the project and underlines some of the main conclusions reached. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hiruta, Mie; Johnson, Gannon [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Rostamian, Maziar, E-mail: mrostamian@asme.org [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Potirniche, Gabriel P. [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States); Ougouag, Abderrafi M. [Idaho National Laboratory, 2525 N Fremont Avenue, Idaho Falls, ID 83401 (United States); Bertino, Massimo; Franzel, Louis [Department of Physics, Virginia Commonwealth University, Richmond, VA 23284 (United States); Tokuhiro, Akira [Department of Mechanical Engineering, University of Idaho, 1776 Science Center Drive, Idaho Falls, ID 83401 (United States)

    2013-10-15

    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.

  16. The Canadian initiative to host the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Dautovich, D.P.; James, R.A.

    1995-01-01

    At the time of the conference, the Canadian Nuclear Fuels Technology Project was making an innovative proposal whereby Ontario Hydro would provide space at its Darlington or Bruce sites as potential sites for the ITER project. An economic impact analysis, conducted by Ernst and Young, showed the potential economic benefits to Canada; other benefits could rather be considered to be scientific and technological benefits. A stable electrical supply grid, existing waste management infrastructure, an abundance of cheap power, and a skilled workforce, made Canada an attractive prospect. ITER, whatever its location, would require all of Ontario Hydro's tritium. Canada was attractive as a neutral siting alternative, and had gained early Russian support

  17. 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)

    Aboudheir, Ahmed; Akande, Abayomi; Idem, Raphael

    2006-01-01

    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 (CO 2 ) emissions resulting from the combustion of these fossil fuels currently are estimated to account for three-fourth of human-caused CO 2 emissions worldwide. Greenhouse gas emission, including CO 2 , should be limited, as recommended at the Kyoto Conference, Japan, in December 1997. In this regard, hydrogen (H 2 ) has a significant future potential as an alternative fuel that can solve the problems of CO 2 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 H 2 and CO 2 , the latter of which could be separated from H 2 by membrane technology. This provides for CO 2 capture for eventual storage or destruction. In the case of using crude ethanol, this will result in negative CO 2 , 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

  18. Activities of research-reactor-technology project in FNCA from FY2005 to FY2007. Sharing neutronics calculation technique for core management and utilization of research reactors

    International Nuclear Information System (INIS)

    2010-07-01

    RRT project (Research-Reactor-Technology Project) was carried out with the theme of 'sharing neutronics calculation technique for core management and utilization of research reactors' in the framework of FNCA (Forum for Nuclear Cooperation in Asia) from FY2005 to FY2007. The objective of the project was to improve and equalize the level of neutronics calculation technique for the reactor core management among participating countries to assure the safe and stable operation of research reactors and the promotion of the effective utilization. Neutronics calculation codes, namely SRAC code system and MVP code, were adopted as common codes. Participating countries succeeded in applying the common codes to analyzing the core of each domestic research reactor. Some participating countries succeeded in applying the common codes to analyzing for utilization of own research reactors. Activities of RRT project have improved and equalized the level of neutronics calculation technique among participating countries. (author)

  19. The reactor safety study of experimental multi-purpose VHTR design

    International Nuclear Information System (INIS)

    Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

    1981-01-01

    Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

  20. Experimental data available for radiation damage modelling in reactor materials

    International Nuclear Information System (INIS)

    Wollenberger, H.

    Radiation damage modelling requires rate constants for production, annihilation and trapping of defects. The literature is reviewed with respect to experimental determination of such constants. Useful quantitative information exists only for Cu and Al. Special emphasis is given to the temperature dependence of the rate constants

  1. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    OpenAIRE

    Zheng, Bin; Liu, Yongqi; Liu, Ruixiang; Meng, Jian; Mao, Mingming

    2015-01-01

    This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h) and catalytic oxidation bed average temperature (20°C to 560°C) within the preheated catalytic oxidation reactor. The pressure drop and res...

  2. Inverse kinetics technique for reactor shutdown measurement: an experimental assessment. [AGR

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, T. A.; McDonald, D.

    1975-09-15

    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. A description is given of an experimental assessment of the technique by investigating known transients created by control rod movements on a small experimental reactor, (2m high, 1m radius). 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.

  3. Cephalopod Experimental Projected Habitat (CEPH: Virtual Reality for Underwater Organisms

    Directory of Open Access Journals (Sweden)

    Noam Josef

    2018-03-01

    Full Text Available Cephalopods' visually driven, dynamic, and diverse skin display makes them a key animal model in sensory ethology and camouflage research. Development of novel methods is critically important in order to monitor and objectively quantify cephalopod behavior. In this work, the development of Cephalopod Experimental Projected Habitat (CEPH is described. This newly developed experimental design bridges computational and ethological sciences, providing a visually controlled arena which requires limited physical space and minimal previous technical background. Created from relatively inexpensive and readily available materials, the experimental apparatus utilizes reflected light which closely resembles natural settings. Preliminary results suggest the experimental design reproducibly challenges marine organisms with visually dynamic surroundings, including videos of prey and predator. This new approach should offer new avenues for marine organism sensory research and may serve researchers from various fields.

  4. 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

  5. Quality assurance in the project of RECH-2 research reactor

    International Nuclear Information System (INIS)

    Goycolea Donoso, C.; Nino de Zepeda Schele, A.

    1989-01-01

    The implantation of a Quality Assurance Program for the design, supply, construction, installation, and testing of the RECH-2 research reactor, is described in this paper. The obtained results, demonstrate that a Quality Assurance Program constitutes a suitable mean to assure that the installation complies with the safety and reliability requirements. (author)

  6. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  7. The Dragon project and high temperature reactor (HTR position)

    International Nuclear Information System (INIS)

    Shepherd, L.

    1981-01-01

    After introduction describing the initiation of HTR work at AERE and in West Germany and the USA, the subject is discussed in detail under the headings: the Dragon Reactor Experiment (design and objectives); fuel elements and graphite (description of cooperative research programmes; development of coated fuel particles); helium technology; other Dragon activities. (U.K.)

  8. Clinch River Breeder Reactor Plant Project: construction schedule

    International Nuclear Information System (INIS)

    Purcell, W.J.; Martin, E.M.; Shivley, J.M.

    1982-01-01

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule

  9. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  10. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

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

    International Nuclear Information System (INIS)

    Miki, Nobuharu; Iida, Fumio; Suzuki, Shohei; Wachi, Yoshihiro; Toyoda, Katsuyoshi; Hashizume, Takashi; Konno, Masayuki.

    1987-09-01

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

  12. Experimental study of defect power reactor fuel. Final report

    International Nuclear Information System (INIS)

    Forsyth, R.S.; Jonsson, T.

    1982-01-01

    Two BWR fuel rods, one intact and one defect, with the same manufacturing and irradiation data have been examined in a comparative study. The defect rod has been irradiated in a defect condition during approximately one reactor cycle and has consequently some secondary defects. The defect rod has two penetrating defects at a distance of about 1.5 meters from each other. Comparison with the intact rod shows a large Cs loss from the defect rod, especially between the cladding defects, where the loss is measured to about 30 %. The leachibility in deionized water is higher for Cs, U and Cm for fuel from the defect rod. The leaching results are more complex for Sr-90, Pu and Am. The fuel in the defect rod has undergone a change of structure with gain growth and formation of oriented fuel structure. The cladding of the defect rod is hydrided locally in some parts of the lower part of the rod and furthermore over a more extended region near the end of the rod. (Authors)

  13. The Canadian initiative to host the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dautovich, D P; James, R A [Canadian Fusion Fuels Technology Project, Mississauga, ON (Canada)

    1996-12-31

    At the time of the conference, the Canadian Nuclear Fuels Technology Project was making an innovative proposal whereby Ontario Hydro would provide space at its Darlington or Bruce sites as potential sites for the ITER project. An economic impact analysis, conducted by Ernst and Young, showed the potential economic benefits to Canada; other benefits could rather be considered to be scientific and technological benefits. A stable electrical supply grid, existing waste management infrastructure, an abundance of cheap power, and a skilled workforce, made Canada an attractive prospect. ITER, whatever its location, would require all of Ontario Hydro`s tritium. Canada was attractive as a neutral siting alternative, and had gained early Russian support.

  14. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2014-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the suing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer is implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer is produced an inverse buoyant force make the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow - open pool research reactor (with a power greater than 20 M watt) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against Gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability

  15. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2015-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the swing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer produced an inverse buoyant force making the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow-open pool research reactor (with a power greater than 20 Mw) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability conditions from

  16. Experimental research on pressurized water reactor(PWR) safety

    International Nuclear Information System (INIS)

    Kim, Dong Su; Chae, Sung Ki; Chang, Won Pyo

    1991-12-01

    The objective of this research is to analyze the experimental results already performed in BETHSY facility of CEA France and to establish essential technologies for the future implementation of both such an experiment and computer code assessment, which are not undergoing in Korea so far. The contents of the present study are divided into 2 categories; namely, analysis of the BETHSY experimental data received from CEA, and CATHARE computer code simulation for the selected experiments, i.e. 'Natural Circulation(Test 4.3a)' and '2 Cold Leg Break'. The later studies are performed under the aims of CATHARE assessment as well as qualification of KOSAC code developing at KAERI, which is the subject in the next year and will concern an adequacy of KOSAC for the prediction of low flow natural circulation and a small break transients. (Author)

  17. Experimental and theoretical studies on hydrogenation of olefins in multiphase fixed bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Battsengel, B.; Datsevitch, L.; Jess, A. [Bayreuth Univ. (Germany). Dept. of Chemical Engineering

    2003-07-01

    Multi phase reactors like trickle bed systems are frequently used for gas-liquid reactions. In general, they have complex mass and heat transfer characteristics; scale-up is therefore difficult. The present work focuses on the role of mass transfer on the effective reaction rate, taking catalytic octene hydrogenation as a model reaction. The reaction rate in a trickle bed reactor is by a factor of about 20 smaller than (theoretically) in the absence of any mass transfer limitations. Based on the experimental results, the so-called pre-saturation concept is presented, where only the liquid saturated with hydrogen is fed into the reactor. The effective reaction rate in this two phase system (liquid and solid cat.) is equal or even higher than in a trickle bed reactor. Scale-up problems do not occur, and the pre-saturation concept has also other advantages (lower energy consumption), as discussed in detail in this paper. (orig.)

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

    International Nuclear Information System (INIS)

    Roddy, J.W.; Mailen, J.C.

    1987-12-01

    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

  19. Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

    2013-05-15

    Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

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

    International Nuclear Information System (INIS)

    Kerveno, Maelle; Dupuis, Marc; Bauge, E.; Hilaire, S.; Romain, P.; Morillon, B.; Delaroche, J.P.; Dupuis, M.; Peru, S.; Belier, G.; Bonnet, T.; Laborie, J.M.; Laurent, B.; Ledoux, X.; Varignon, C.; Meot, V.; Bernard, David; Capote, Roberto; Kawano, T.; Bond, E.; Vieira, D.J.; Wilhelmy, J.B.; Raynal, J.; Plompen, Arjan J.M.; Drohe, J.C.; Nankov, N.; Nyman, M.; Rouki, C.; Bacquias, A.; Dessagne, Ph.; Henning, G.; Karam, H.; Kerveno, M.; Rudolf, G.; Thiry, J.C.; Borcea, C.; Negret, A.; Stanoiu, M.; Bucurescu, D.; Deleanu, D.; Filipescu, D.; Ghita, D.; Glodariu, T.; Marginean, N.; Marginean, R.; Mihai, C.; Olacel, A.; Pascu, S.; Sava, T.; Stroe, L.; Goriely, S.; Pavlik, A.; Jericha, E.; Ledoux, X.; Becker, J.A.; Macri, R.; Authier, N.; Hyneck, D.; Jansen, Y.; Legendre, J.; Jacquet, X.; Gunsing, Frank; Henning, Greg

    2014-03-01

    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

  1. Smart grid development and households in experimental projects

    DEFF Research Database (Denmark)

    Hansen, Meiken

    to the electricity grids and call for the development of smart grids. The Danish Smart Grid Strategy states that ‘flexible electricity consumption’ is the main purpose of smart grids in Denmark, envisioning that future consumers will have flexible consumption of electricity. Thus, they are expected to respond...... to the supply side and consume energy when it is available. The goal of this thesis is to investigate how household consumers are integrated in smart grid development activities. More specifically, it focuses on household consumers, as they are represented in experimental projects in the smart grid area...... been little research on the area in Danish smart grid experimental projects. Overall, the consumers are expected, to some extent, to provide flexibility by changing their energy-consuming practices because of economic incentives by means of manual or automated control of devices. Moreover, the Danish...

  2. Recommendations for a cryogenic system for ITER [International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Slack, D.S.

    1989-01-01

    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: Only time-proven components are used. The system obtains a high efficiency without use of cold pumps or other developmental components. High reliability is achieved by paralleling compressors and expanders and by using adequate isolation valving. The problem of load fluctuations is solved by a simple load-leveling device. 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 magnet quench. 4 refs., 4 figs., 3 tabs

  3. Design considerations for ITER [International Thermonuclear Experimental Reactor] toroidal field coils

    International Nuclear Information System (INIS)

    Kalsi, S.S.; Lousteau, D.C.; Miller, J.R.

    1987-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Europe, Japan, the Union of Soviet Socialist Republics (USSR), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb/sub 3/Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the US TIBER concept. The nuclear heating generated during operation is removed from the windings by helium flowing through the conductor. The heat in the coil case is removed through a separate cooling circuit operating at approximately 20 K. Manifold concepts are presented for the complete coil cooling system. Also included are concepts for the coil structural arrangement. The effects of in-plane and out-of-plane loads are included in the design considerations for the windings and case. Concepts are presented for reacting these loads with a minimum amount of additional structural material. Concepts discussed in this paper could be considered for the ITER TF coils. 6 refs., 5 figs., 1 tab

  4. Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

    International Nuclear Information System (INIS)

    Devan, K.; Bachchan, Abhitab; Riyas, A.; Sathiyasheela, T.; Mohanakrishnan, P.; Chetal, S.C.

    2011-01-01

    Highlights: → In this study we made physics designs of experimental metal fast reactor cores. → Aim is for full-scale demonstration of fuel assemblies in a commercial power reactor. → Minimum power with adequate safety is considered. → In addition, fuel sustainability is also considered in the design. → Sodium bonded U-Pu-6%Zr and mechanically bonded U-Pu alloys are used. - Abstract: Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.

  5. Balance of plant improvements for future reactor projects

    International Nuclear Information System (INIS)

    Hollingshaus, H.

    1987-01-01

    Many studies have shown that improvements in portions of the plant other than the reactor systems can yield large cost savings during the design, construction, and operation of future reactor power plants. This portion is defined as the Balance of Plant which includes virtually everything except the equipment furnished by the Nuclear Steam Supply System manufacturer. It normally includes the erection of the entire plant including the NSSS. Cost of BOP equipment, engineering and construction work is therefore most of the cost of the plant. Improvements in the BOP have been identified that will substantially reduce nuclear plant cost and construction time while at the same time increasing availability and operability and improving safety. Improvements achieved through standardizatoin, simplification, three-dimensional (3D) computer-aided design, modular construction, innovative construction techniques, and applications for Artificial Intelligence Systems are described. (author)

  6. Project of a binary breeder reactor and its inherent safety

    International Nuclear Information System (INIS)

    Nascimento, J.A. do; Dias, A.F.; Ishiguro, Y.

    1983-01-01

    A core layout for the binary breeder reactor (BBR) is developed based on the results of preliminary burnup calculations. The apparent breeding ratio, in the U 233 /Th fueled inner core, is low due to the accumulation of Pa-233 in the first few months of operation. The loss of reactivity during this time is around 3%. The BBR requires more reactivity control than Pu/U-fueled LMFBRs and the core layout developed has 19 control rod assemblies in the inner core. Three aspects related to the inherent safety of the Binary Breeder Reactor have been studied: the radial distribution of the sodium-void reactivity zone-wise Doppler reactivity and the fractions of delayed neutrons. The results show excellent characteristics for the BRB safety. (Author) [pt

  7. The risks of the Taiwan research reactor spent fuel project

    International Nuclear Information System (INIS)

    1991-06-01

    The proposed action is to transport up to 118 spent fuel rods, to include canned spent fuel rod particulates immobilized on filters, from a research reactor in Taiwan by sea to Hampton Roads, Virginia, and then overland by truck to the Receiving Basin for Offsite Fuels at the Savannah River Site (SRS). At SRS, the spent fuel will be reprocessed to recover uranium and plutonium. 55 refs., 8 tabs

  8. Acquisition of reactor experimental data; Akviziter reaktorskih eksperimentalnih podataka

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Tasic, A [Institut za nuklearne nauke ' Boris Kidric' , Vinca, Belgrade (Yugoslavia)

    1966-07-01

    This paper include the analysis of possible experiments and relevant experimental devices for detection, registering and analysis of inducing and response signals. It contains a concept of our system for detection and registering of data, which i appropriate for our research program. Non-typical details of certain acquisition systems are described as well. U ovom radu se analiziraju moguci eksperimenti i odgovarajuci eksperimentalni uredjaji za detekciju, registraciju i analizu signala pobude i odziva. Dalje se iznosi koncepcija naseg sistema za detekciju i registraciju podataka pogodnog za nas program istrazivanja. Netipicni detalji pojedinih kola akvizitera takodje se iznose u radu (author)

  9. Experimental estimations of the kinetics parameters of the IBR-2M reactor by stochastic noises

    International Nuclear Information System (INIS)

    Pepelyshev, Yu.N.; Tajybov, L.A.; Garibov, A.A.; Mekhtieva, R.N.

    2012-01-01

    Experimental investigations of stochastic fluctuations of pulse energy of the IBR-2M reactor have been carried out which allowed us to obtain some of the parameters of the reactor kinetics. At different levels of average power a sequence of values of pulse energy was recorded with the calculation of the distribution parameters. An ionization chamber with boron installed near the active zone was used as a neutron detector. The research results allowed us to estimate the average lifetime of prompt neutrons τ = (6.53±0.2)·10 -8 s, absolute power of the reactor and intensity of the source of spontaneous neutrons S sp ≤(6.72±0.12)·10 6 s -1 . It was shown that the experimental results are close to the calculated ones

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

    International Nuclear Information System (INIS)

    Cummins, J.D.

    1968-04-01

    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)

  11. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  12. Development of observation techniques in reactor vessel of experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Takamatsu, Misao; Imaizumi, Kazuyuki; Nagai, Akinori; Sekine, Takashi; Maeda, Yukimoto

    2010-01-01

    In-Vessel Observations (IVO) techniques for Sodium cooled Fast Reactors (SFRs) are important in confirming its safety and integrity. And several IVO equipments for an SFR are developed. However, in order to secure the reliability of IVO techniques, it was necessary to demonstrate the performance under the actual reactor environment with high temperature, high radiation dose and remained sodium. During the investigation of an incident that occurred with Joyo, IVO using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS). A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1 mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8 mm in thickness could be discriminated. The performance of IVO equipments under the actual reactor environment was successfully confirmed. And the results provided useful information on incident investigations. In addition, fundamental findings and the experience gained during this study, which included the design of equipment, operating procedures, resolution, lighting adjustments, photograph composition and the durability of the RRF under radiation exposure, provided valuable insights into further improvements and verifications for IVO techniques to

  13. A summary of lessons learned activities conducted at the OECD Halden Reactor Project

    International Nuclear Information System (INIS)

    Hallbert, B.P.

    1997-01-01

    A series of lessons learned studies have been conducted at the OECD Halden Reactor Project. The purpose of these lessons learned reports are to summarize knowledge and experience gained across a number of research project. This paper presents a summary of main issues addressed in four of these lessons learned projects. These are concerned with software development and quality assurance, software reliability, methods for test and evaluation of developed systems, and the evaluation of system design features

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

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

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

    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

  17. Experimental determination of lattice parameters for 2% enriched uranium heavy water reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N; Takac, S; Markovic, H; Bosevski, T [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Systematic measurements of the buckling, infinite multiplication factor and the thermal utilization factor were made on a series of lattices for 2% enriched uranium tubular fuel elements in heavy water. This work represents the first phase of experimental verification of standard theoretical methods used for the determination of reactor parameters.

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

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

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

  19. The gas turbine modular helium reactor. An international project to develop a safe, efficient, flexible product

    International Nuclear Information System (INIS)

    Silberstein, A.J.

    1998-01-01

    As originally scheduled, the Conceptual Design Report of the 600 Mwt Gas Turbine Modular Helium Reactor has been issued in October 1997 by OKBM in Nizhny Novgorod, a keystone Russian Engineering Institute fully involved in the realization of this International Project. The plutonium burning, graphite moderated helium cooled reactor design results from the work done on the basis of General Atomics original concept combined with the goal of optimizing safety power and efficiency with multi contributions in specific fields from the Russian organizations: MINATOM, OKBM, VNIINM, Lutch, Kurchatov Institute, Seversk Chemical Combinat, Fuji Electric and FRAMATOME. The objective to concentrate the engineering work in Russia has met a full success due principally to the quality and experience of the people, to the international support and to the progressive integration of new techniques of communication, of project management culture and utilization of modern computerized design tools and methods. To day the best international standard of quality is reached in the engineering activity and expected to stay at this level for future developments, when including experimental facilities operation and components manufacturing activities, thanks to the diffusion of the common culture, acquired by the main actors during the conceptual design phase, that will be exported to Russian third parties. At this stage we are planning to start design verification and sensitive components and systems qualification, with the same original actors. The European Commission has already shown some significant interest through the MICHELANGELO Initiative in supporting the HTR concepts assessment and identification of the R and D needs. We are looking forward for further support from the International Community and particularly from European Institutions in the frame of the 5th PCRD to pursue the GT MHR R and D program. Furthermore we are looking for funding the building of a prototype in Russia

  20. Vietnam Project For Production Of Radioactive Beam Based On ISOL Technique With The Dalat Reactor

    International Nuclear Information System (INIS)

    Le Hong Khiem; Phan Viet Cuong; Fadi Ibrahim

    2011-01-01

    The presence in Vietnam of Dalat nuclear reactor dedicated to fundamental studies is a unique opportunity to produce Radioactive Ion (RI) Beams with the fission of a 235 U induced by the thermal neutrons produced by the reactor. We propose to produce RI beams at the Dalat nuclear reactor using ISOL (Isotope Separation On-Line) technique. This project should be a unique opportunity for Vietnamese nuclear physics community to use its own facilities to produce RI beams for studying nuclear physics at an international level. (author)

  1. Experimental verification of creep analyses for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Abe, H.; Ohnuma, H.

    1977-01-01

    The authors proposed a new method of creep analysis based on the theory of strain hardening, which assumes that accumulated creep at a given time influences the creep after that. This method was applied to calculate step-by-step the behaviors of uniaxial creep of concrete under variable temperatures and stresses, creep in reinforced concrete specimens and the behaviors of prestressed concrete beams under themal gradients. The experimental and calculated results agreed fairly well. Further, this method was incorporated in the finite element creep analysis for the prestressed concrete hollow cylinder and the full scale model. The calculated strain changes with time pursued closely those obtained by experiments. The above led to the conclusion that from the viewpoint of both accuracy and computation time the strain hardening method proposed by the authors may be judged advantageous for practical usages

  2. Project management system for the decommissioning of research reactors

    International Nuclear Information System (INIS)

    Park, J. H.

    2006-01-01

    KAERI has developed a computer information system, named DECOMMIS, for the project management with the increased effectiveness of the decommissioning projects and the record keeping for a next decommissioning project. The management system consists of three parts, code management system, data input system (DDIS) and data processing and output system (DDPS). Through the DDIS, the data can be directly inputted at sites and the system can play roles of daily work reports to minimize the time gap between the dismantling activities and the evaluation of the data for project management. The DDPS provides useful information to the staff for more effective project management and this information include several fields, such as project progress management, man power management, waste management, radiation dose of workers and so on. It is expected that the system would enable to maintain the decommissioning data, to prepare the source data for the R and D for development of planning tools and to give information to the staff for the decision on the progress of the projects. In this paper, the overall system will be briefly explained and several examples of the utilization, focused on the waste and manpower control, for the project management will be introduced

  3. On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho

    2007-01-01

    In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

  4. Seismic isolation of lead-cooled reactors: The European project SILER

    International Nuclear Information System (INIS)

    Forni, Massimo; Poggianti, Alessandro; Scipinotti, Riccardo; Dusi, Alberto; Manzoni, Elena

    2014-01-01

    SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the 7th Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the 6th Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

  5. Approaches to safety, environment and regulatory approval for the International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

    Saji, G.; Bartels, H.W.; Chuyanov, V.; Holland, D.; Kashirski, A.V.; Morozov, S.I.; Piet, S.J.; Poucet, A.; Raeder, J.; Rebut, P.H.; Topilski, L.N.

    1995-01-01

    International Thermonuclear Experimental Reactor (ITER) Engineering Design Activities (EDA) in safety and environment are approaching the point where conceptual safety design, topic studies and research will give way to project oriented engineering design activities. The Joint Central Team (JCT) is promoting safety design and analysis necessary for siting and regulatory approval. Scoping studies are underway at the general level, in terms of laying out the safety and environmental design framework for ITER. ITER must follow the nuclear regulations of the host country as the future construction site of ITER. That is, regulatory approval is required before construction of ITER. Thus, during the EDA, some preparations are necessary for the future application for regulatory approval. Notwithstanding the future host country's jurisdictional framework of nuclear regulations, the primary responsibility for safety and reliability of ITER rests with the legally responsible body which will operate ITER. Since scientific utilization of ITER and protection of the large investment depends on safe and reliable operation of ITER, we are highly motivated to achieve maximum levels of operability, maintainability, and safety. ITER will be the first fusion facility in which overall 'nuclear safety' provisions need to be integrated into the facility. For example, it will be the first fusion facility with significant decay heat and structural radiational damage. Since ITER is an experimental facility, it is also important that necessary experiments can be performed within some safety design limits without requiring extensive regulatory procedures. ITER will be designed with such a robust safety envelope compatible with the fusion power and the energy inventories. The basic approach to safety will be realized by 'defense-in-depth'. (orig.)

  6. The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

    International Nuclear Information System (INIS)

    Gone, J.K.; Huang, Y.H.

    2001-01-01

    The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E 14 n/cm 2 sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

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

    Science.gov (United States)

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

    2017-01-01

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

  8. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  9. Decommissioning of the ASTRA research reactor - planning, executing and summarizing the project

    International Nuclear Information System (INIS)

    Meyer, F.

    2010-01-01

    The decommissioning of the ASTRA research reactor at the Austrian Research Centres Seibersdorf was described within three technical papers already released in Nuclear Technology and Radiation Protection throughout the years 2003, 2006, and 2008. Following a suggestion from IAEA the project was investigated well after the files were closed regarding rather administrative than technical matters starting with the project mission, explaining the project structure and identifying the key factors and the key performance indicators. The continuous documentary and reporting system as implemented to fulfil the informational needs of stakeholders, management, and project staff alike is described. Finally the project is summarized in relationship to the performance indicators. (author)

  10. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    2001-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

  11. International Thermonuclear Experimental Reactor (ITER) plant layout and site services

    International Nuclear Information System (INIS)

    Chuyanov, V.

    1999-01-01

    The ITER site has not been determined at this time. Nevertheless, to develop a construction plan and a cost estimate, it is necessary to have a detailed layout of the buildings, structures, and outdoor equipment integrated with the balance of plant service systems prototypical of large fusion power plants. These services include electric power for magnet feeds and plasma heating systems, cryogenic and conventional cooling systems, compressed air, gas supplies, de-mineralized water, steam, and drainage. Nuclear grade facilities are provided to handle tritium fuel and activated waste, as well as to prevent radioactive exposure of either the workers or the public. To avoid interference between services of different types and for efficient arrangement of buildings, structures, and equipment within the site area, a plan was developed which segregated different classes of services to four quadrants surrounding the tokamak building, placed at the approximate geographic center of the site. Location of the twenty-seven buildings on the generic site was selected to meet all design requirements at minimum total project cost. A similar approach has been used to determine the location of services above, at, and below grade. The generic site plan can be adapted to the site selected for ITER without significant changes to the buildings or equipment. Some rearrangements may be required by site topography resulting primarily in changes to the length of services that link the buildings and equipment. (author)

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

    International Nuclear Information System (INIS)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C.

    2017-01-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO_2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D_2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α"M_T(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  14. Experimental study on reactivity measurement in thermal reactor by polarity correlation method

    International Nuclear Information System (INIS)

    Yasuda, Hideshi

    1977-11-01

    Experimental study on the polarity correlation method for measuring the reactivity of a thermal reactor, especially the one possessing long prompt neutron lifetime such as graphite on heavy water moderated core, is reported. The techniques of reactor kinetics experiment are briefly reviewed, which are classified in two groups, one characterized by artificial disturbance to a reactor and the other by natural fluctuation inherent in a reactor. The fluctuation phenomena of neutron count rate are explained using F. de Hoffman's stochastic method, and correlation functions for the neutron count rate fluctuation are shown. The experimental results by polarity correlation method applied to the β/l measurements in both graphite-moderated SHE core and light water-moderated JMTRC and JRR-4 cores, and also to the measurement of SHE shut down reactivity margin are presented. The measured values were in good agreement with those by a pulsed neutron method in the reactivity range from critical to -12 dollars. The conditional polarity correlation experiments in SHE at -20 cent and -100 cent are demonstrated. The prompt neutron decay constants agreed with those obtained by the polarity correlation experiments. The results of experiments measuring large negative reactivity of -52 dollars of SHE by pulsed neutron, rod drop and source multiplication methods are given. Also it is concluded that the polarity and conditional polarity correlation methods are sufficiently applicable to noise analysis of a low power thermal reactor with long prompt neutron lifetime. (Nakai, Y.)

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

    International Nuclear Information System (INIS)

    Paraschiv, Marius; Turnbull, J.A.

    2002-01-01

    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

  16. 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

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

    International Nuclear Information System (INIS)

    He Jie; Xu Xianhong

    2010-01-01

    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.

  18. Mechatronics of fuel handling mechanism for fast experimental reactor 'Joyo'

    Energy Technology Data Exchange (ETDEWEB)

    Fujiwara, Akikazu (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center)

    1984-01-01

    The outline of the fast experimental reactor ''Joyo'' is introduced, and the fuel handling mechanism peculiar to fast reactors is described. The objectives of the construction of Joyo are to obtain the techniques for the design, construction, manufacture, installation, operation and maintenance of sodium-cooled fast reactors independently, and to use it as an irradiation facility for the development of fuel and materials for fast breeder reactors. At present, the reactor is operated at 100 MW maximum thermal output for the second objective. Since liquid sodium is used as the coolant, the atmosphere of the fuel handling course changes such as liquid sodium at 250 deg C, argon gas at 200 deg C and water, in addition, the spent fuel taken out has the decay heat of 2.1 kW at maximum. The fuel handling works in the reactor and fuel transfer works, and the fuel handling mechanism of a fuel exchanger and that of a cask car for fuel handling are described. Relay sequence control system is used for the fuel handling mechanism of Joyo.

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

    Science.gov (United States)

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

    2018-01-01

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

  20. Construction of fast experimental reactor 'Joyo' from start of construction to criticality

    International Nuclear Information System (INIS)

    Sakata, Hajime

    1977-01-01

    The fast experimental reactor ''Joyo'' is a sodium-cooled, fast neutron reactor using mixed oxide of uranium and plutonium, the first in Japan. The purposes of its construction are to experience and solve the various technical problems expected in the constructions of the prototype reactor ''Monju'' and future practical reactors, and to use as the irradiation facility for developing the fuel and material for fast breeder reactors in Japan after the completion. The construction finished by the end of 1974, and the synthetic functional test was carried out for about two years thereafter. The whole installation was handed over to PNC on March 8, 1977. The reactor attained the criticality on April 24, 1977. The outline of the construction works is described. ''Guidance to the structural design of sodium machinery for Joyo'' was compiled, and the analysis was made according to it. Moreover, various inspection standards regarding welding, electrical machinery, fuel and others were made. The revision of the design for improving the safety and performance was made during the construction at all times. The synthetic functional test was carried out for about two years on 266 items, and subsequently, the criticality test was completed satisfactorily. (Kako, I.)

  1. 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.

  2. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    International Nuclear Information System (INIS)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

  3. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  5. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  6. Response tree evaluation: experimental assessment of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1985-09-01

    The United States Nuclear Regulatory Commission (USNRC) sponsored a project performed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) to evaluate different display concepts for use in nuclear reactor control rooms. Included in this project was the evaluation of the response tree computer-based decision aid and its associated displays. The response tree evaluation task was deisgned to (1) assess the merit of the response tree decision aid and (2) develop a technical basis for recommendations, guidelines, and criteria for the design and evaluation of computerized decision aids for use in reactor control rooms. Two major experiments have been conducted to evaluate the response tree system. This report emphasizes the conduct and results of the second experiment. An enhanced version of the response tree system, known as the automated response tree system, was used in a controlled experiment using trained reactor operators as test subjects. This report discusses the automated response tree system, the design of the evaluation experiment, and the quantitative results of the experiment. The results of the experiment are compared to the results of the previous experiment to provide an integrated perspective of the response tree evaluation project. In addition, a subjective assessment of the results addresses the implications for the use of advanced, ''intelligent'' decision aids in the reactor control room

  7. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

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

    International Nuclear Information System (INIS)

    Maas, L.; Beuter, A.; Seropian, C.

    2010-01-01

    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

  9. Quality assurance in technology development for The Clinch River Breeder Reactor Plant Project

    International Nuclear Information System (INIS)

    Anderson, J.W.

    1980-01-01

    The Clinch River Breeder Reactor Plant Project is the nation's first large-scale demonstration of the Liquid Metal Fast Breeder Reactor (LMFBR) concept. The Project has established an overall program of plans and actions to assure that the plant will perform as required. The program has been established and is being implemented in accordance with Department of Energy Standard RDT F 2-2. It is being applied to all parts of the plant, including the development of technology supporting its design and licensing activity. A discussion of the program as it is applied to development is presented

  10. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  11. Experimental and analytical studies on soil-structure interaction behavior of nuclear reactor building

    International Nuclear Information System (INIS)

    Tsushima, Y.

    1978-01-01

    The purpose of this study is to estimate damping effects due to soil-structure interaction by the dissipation of vibrational energy to the ground through the foundation in a building with a short fundamental period such as a nuclear reactor building. The author performed experimental and analytical studies on the vibrational characteristics of model steel structures ranging from one to four stories high erected on the rigid base and located on soil, which are simulated from the vibrational characteristics of a prototype reactor building: the former study is to obtain damping effects due to inner friction of steel frames and the latter to obtain radiation damping effects due to soil-structure interaction. The author also touches upon the results of experiments performed on a BWR-type reactor building in 1974, which showed damping ratios higher than 20% of those in fundamental modes. Then the author attempts to estimate the damping effects of the reactor building by his own method proposed in the report. Through these studies the author finally concludes that the experimental damping effects are remarkable in the lower modes by the energy dissipation and the analytical results show a fairly good fit to the experimental ones

  12. Experimental Investigation of Flow Resistance in a Coal Mine Ventilation Air Methane Preheated Catalytic Oxidation Reactor

    Directory of Open Access Journals (Sweden)

    Bin Zheng

    2015-01-01

    Full Text Available This paper reports the results of experimental investigation of flow resistance in a coal mine ventilation air methane preheated catalytic oxidation reactor. The experimental system was installed at the Energy Research Institute of Shandong University of Technology. The system has been used to investigate the effects of flow rate (200 Nm3/h to 1000 Nm3/h and catalytic oxidation bed average temperature (20°C to 560°C within the preheated catalytic oxidation reactor. The pressure drop and resistance proportion of catalytic oxidation bed, the heat exchanger preheating section, and the heat exchanger flue gas section were measured. In addition, based on a large number of experimental data, the empirical equations of flow resistance are obtained by the least square method. It can also be used in deriving much needed data for preheated catalytic oxidation designs when employed in industry.

  13. Experimental and calculational works on characteristics of the Dalat Nuclear Research Reactor. Second edition

    International Nuclear Information System (INIS)

    Pham Ngoc Khoi; Nguyen Kim Dung

    2016-03-01

    Recognizing the significant value and necessity of publishing the scientific document of experimental and calculational works on the Dalat Nuclear Research Reactor (DNRR) physics and engineering for research, operation, training activities as well as for international scientific exchange, Vietnam Atomic Energy Agency (VAEA) and Vietnam Atomic Energy Institute have completed editing to publish the “Experimental and Calculational Works on Characteristics of THE DALAT NUCLEAR RESEARCH REACTOR” which consists of 26 typical papers representing the most important experimental and calculational results of the DNRR physics and engineering obtained during 30 years of operation and exploitation with the contribution of Vietnamese and former USSR’s experts, especially scientists and engineers working at the Reactor Center of the NRI

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

    International Nuclear Information System (INIS)

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  15. Some considerations on a plasma in the JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Tone, T.; Yamato, H.; Maki, K.

    1976-01-01

    The preliminary analysis of the plasma characteristics for the JAERI tokamak experimental fusion reactor is reported. In order to make the reactor compact, the self-sustaining condition has been removed. Stationary heating by 200 keV neutral deuteron beam to maintain the power balance is applied expecting the power amplification by the TCT effect. The main parameters determined are power output of 100 MW, toroidal field on axis of 6 T, aspect ratio of 4.5 and major radius of 6.75 m. The results of the plasma power balance, fueling by means of the gas blanket scheme, power stabilization with feedback and the start-up are presented

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

    International Nuclear Information System (INIS)

    Adachi, J.; Kakudate, S.; Oka, K.; Seki, M.

    1995-01-01

    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.)

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

    International Nuclear Information System (INIS)

    1983-03-01

    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. [fr

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

    International Nuclear Information System (INIS)

    Roussos, N.

    1982-01-01

    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 [fr

  19. Feasibility of reactivity worth measurements by perturbation method with Caliban and Silene experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, Pierre; Authier, Nicolas [Commissariat a l' Energie Atomique, Centre d' Etudes de Valduc, 21120 Is-Sur-Tille (France)

    2008-07-01

    Reactivity worth measurements of material samples put in the central cavities of nuclear reactors allow to test cross section nuclear databases or to extract information about the critical masses of fissile elements. Such experiments have already been completed on the Caliban and Silene experimental reactors operated by the Criticality and Neutronics Research Laboratory of Valduc (CEA, France) using the perturbation measurement technique. Calculations have been performed to prepare future experiments on new materials, such as light elements, structure materials, fission products or actinides. (authors)

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

    International Nuclear Information System (INIS)

    1985-12-01

    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. The results of the following design items are included: heating/current drive system, plasma position control, power supply, diagnostics, neutronics, blanket test module, repair and maintenance and safety. (author)