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Sample records for cadarache maquette surgeneratic reactor

  1. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

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

  2. ITER at Cadarache

    International Nuclear Information System (INIS)

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

  3. Recent developments in post-irradiation examination techniques for water reactor fuel. Proceedings of a technical committee meeting held in Cadarache, France, 17-21 October 1994

    International Nuclear Information System (INIS)

    At the invitation of the Government of France, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) the IAEA convened a Technical Committee meeting from 14 to 21 October 1994 in Cadarache to discuss recent technical advances and improvements in the field of post-irradiation examination (PIE) of fuel used in nuclear power plants. Fifty participants representing 14 countries attended the meeting and 30 papers were presented and discussed during five technical sessions. Working Groups composed of the session chairmen and authors of papers prepared summaries of each session including conclusions and recommendations for future work. Refs, figs and tabs

  4. The key role of critical mock-up facilities for neutronic physics assessment of advanced reactors: an overview of Cea Cadarache tools

    International Nuclear Information System (INIS)

    The Experimental Physics section of CEA Cadarache operates three critical facilities devoted to neutronic studies of advanced reactors (EOLE, MINERVE and MASURCA) covering a large scope of interests. These include 100% MOX core in ABWR qualification, knowledge improvement of basic nuclear data for heavy nuclides for new options of the fuel cycle - especially the multi-recycling of plutonium - and accelerator-driven systems neutronic behaviour for transmutation studies. The paper describes these facilities, the scientific programmes associated and the progressive improvement of experimental techniques, the aim being to significantly reduce the uncertainties regarding the evaluation of the physical parameters. (authors)

  5. Report transparency and nuclear safety 2007 CEA Cadarache; Rapport transparence et securite nucleaire 2007 CEA Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    This report presents the activities of the CEA Center of Cadarache for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the report discusses the beginning of the RJH reactor construction, the fourth generation reactors research programs, the implementing of la Rotonde the new radioactive wastes management installation, the renovation of the LECA. (A.L.B.)

  6. Cadarache - 20 years of plutonium fuels

    International Nuclear Information System (INIS)

    The qualitative and quantitative evolution of plutonium used is reviewed; more particularly, the isotopic composition of plutonium handled at the plutonium fuel fabrication plant of Cadarache has been considerably changing during these 20 last years. The evolution of fast neutron reactor fuel assemblies explains why the fabrication unit has to. The installations have been modified, as also the fuel assembly fabrication process. Changes can be classified in 3 categories: the transformations related to mass evolution, those related more particularly to the plutonium isotopic composition, finally those related to the waste treatment process. These transformations concerned workshops, apparatus and equipments, the reinforcement of protections, criticality accident prevention, and safeguards

  7. Report transparency and nuclear safety 2007 CEA Cadarache

    International Nuclear Information System (INIS)

    This report presents the activities of the CEA Center of Cadarache for the year 2007. The actions concerning the safety, the radiation protection, the significant events, the release control and the environmental impacts and the wastes stored on the center are discussed. More especially the report discusses the beginning of the RJH reactor construction, the fourth generation reactors research programs, the implementing of la Rotonde the new radioactive wastes management installation, the renovation of the LECA. (A.L.B.)

  8. The IRSN Institute of radiation protection and nuclear safety Cadarache Center

    International Nuclear Information System (INIS)

    The research programs of the IRSN in the Cadarache Center concern the nuclear safety (reactor safety, nuclear fuels behavior during accidents, fires in nuclear installations), the protection and the control of radioactive materials, the human and the environment protection. The programs are presented and discussed. This presentation includes also the regional impact of the Institute at Cadarache (economical impact, relations with the universities and international meetings). (A.L.B.)

  9. Environmental monitoring in the area of the Cadarache Nuclear Studies Centre

    International Nuclear Information System (INIS)

    The Cadarache Centre is part of the French nuclear programme providing for extensive development of research and hence of personnel and installations. It is essentially a centre for testing prototype reactors intended for use in power stations or for ship propulsion. In addition, experimental studies are carried out in connection with reactor structural materials and fuel elements. Various studies related to reactor physics, radiation protection, plant biology, etc. are also undertaken. This article describes the Cadarache site and the monitoring programme to assess the levels of radioactivity released from it.

  10. Contribution of the Maeva mockup for the understanding of the behaviour of reactor containment in accident situations; Apports de la maquette Maeva pour la comprehension du comportement des enceintes de confinement en situation accidentelle

    Energy Technology Data Exchange (ETDEWEB)

    Coudert, L. [Electricite de France (EDF/SEPTEN), 69 - Villeurbanne (France)

    2004-07-01

    The Maeva mockup has been designed to better understand the behaviour of the reactor containment building in accident situations. Maeva has a cylindrical shape, is 5 m high, has a diameter and wall thickness of respectively 16 m and 1.2 m. The concrete composition, the wall thickness and the volume/internal-concrete-surface ratio are identical to those of a N4-type PWR containment in order to avoid scaling problem and get a right simulation of reality. An outside metal wall surrounding the concrete cylinder creates a tight space in which the leaks through the 1.2 m thick wall can be assessed and measured with adequate instrumentation. Other instrumentation is dedicated to temperature and deformation measurement through sensors embedded in the wall thickness. Maeva has been operating since 1998, 5 measurement campaigns have been performed, they have allowed us to check measured and computed data about the containment deformation and to link the appearance rate of cracks with the pressure level inside the containment. Thermal and pressure loads generated during measurement campaigns will limit Maeva operating life. (A.C.)

  11. Enriched uranium workshops towards decommissioning CEA Cadarache

    International Nuclear Information System (INIS)

    The author presents the post operational clean-out following the shut down of the ATUE nuclear facility implemented in the Cadarache research Center (France) and preceding the dismantling operations. ATUE was constructed in 1965, in the support of the fuel cycle industry, to develop the dry process of UF6 to UO2 conversion and the wet process with step of nitric acid solution, solvent purification by exchange columns, ammonium precipitation and hydrogen reduction. The dismantling organization, methods and technology are also presented and discussed. In a last part the author highlights some important points of the project. (A.L.B.)

  12. Radiation protection at the Cadarache research center

    International Nuclear Information System (INIS)

    This article recalls the French law about radiation protection and its evolution due to the implementation of the 2013/59-EURATOM directive that separates the missions of counsel from the more operative missions of the person appointed as 'competent in radiation protection'. The organisation of the radiation protection of the Cadarache research center is presented. The issue of sub-contracting and the respect of an adequate standard of radioprotection is detailed since 2 facilities operated by AREVA are being dismantled on the site. (A.C.)

  13. Remote Participation at JET from DRFC Cadarache

    International Nuclear Information System (INIS)

    The DRFC at the CEA/Cadarache has participated in the EFDA-JET Programme via the Task Agreement for Profile Control Studies in 1999 and via the EFDA-JET agreement in 2000. Several physicists have been seconded to Culham to work JET and many others have participated in the program 'at distance'. This task has necessitated careful consideration of working methods, often with 'ad-hoc' remote working solutions, but has led to a clear identification of the technical needs for Remote Participation (RP) for the new European Fusion Development Agreement (EFDA). We have contributed substantial input into the working groups evaluating the technical needs of RP

  14. Integral needs for MOX powders: state of the art at Cea Cadarache on MOX fuel experiments

    International Nuclear Information System (INIS)

    Several experimental programmes have been conducted at CEA connected with the use of MOX in commercial PWR or innovative concepts, such as high conversion light water reactors (HCLWRs) and at various moderation ratios. Measurements of integral parameters of interest were performed using fission chambers and gamma-scanning techniques, or by oscillation techniques. This paper presents an overview of the experimental programmes performed at the EOLE and MINERVE facilities located at CEA Cadarache. (author)

  15. [Concorso di progettazione Piazza Verdi - La Spezia. Fotografia della maquette2

    OpenAIRE

    Vitone, Luca; Boehm, Frank

    2016-01-01

    2.JPG (file jpg, 31,82x21,17 cm, 72 dpi) Fotografia della maquette in compensato del progetto per la piazza, dettaglio. 2.JPG (file jpg, 31,82x21,17 cm, 72 dpi) Image of the plywood scale model of the square, detail.

  16. [Concorso di progettazione Piazza Verdi - La Spezia. Fotografia della maquette5

    OpenAIRE

    Vitone, Luca; Boehm, Frank

    2016-01-01

    5.JPG (file jpg, 31,82x21,17 cm, 72 dpi)otografia della maquette in compensato del progetto per la piazza, dettaglio. 5.JPG (file jpg, 31,82x21,17 cm, 72 dpi) Image of the plywood scale model of the square, detail.

  17. The Cadarache 1 MV porcelain SINGAP bushing

    International Nuclear Information System (INIS)

    The 1 MV bushing is considered as one of the most technically challenging components of the ITER neutral beam injectors, necessitating the design and manufacture of large ceramic insulators beyond the range of current industrial production. The ITER radiation environment and the fact that the bushing forms part of the tritium barrier demand a component of great reliability and safety. An R and D task has been launched for the fabrication of a prototype bushing based on the SINgle GAP, SINGle APerture (SINGAP) 1 MeV accelerator being developed in the EU. We present the design of this prototype, which is to be tested at Cadarache, and compare it with the SINGAP bushing proposed for the ITER injectors

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

    International Nuclear Information System (INIS)

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

  19. TJ-II operation tracking from Cadarache

    International Nuclear Information System (INIS)

    The TJ-II remote participation system was designed to follow the TJ-II discharge production, even allowing the physicist in charge of operation to be in a remote location. The system has been based on both web servers and Java technology. These elements were chosen due to its open character, security properties, platform independence and technological maturity. Web pages and Java applications permit users to access experimental systems, data servers and the operation logbook. Security resources are provided by the PAPI system, a distributed authentication and authorization system. The TJ-II remote participation tools have allowed us to command and follow the stellarator operation from Cadarache. Over 1,000 digitizer channels and more than 20 diagnostic control systems were remotely available from web pages for monitoring/programming purposes. One Java application provided on-line information about the acquisition status of channels and acquisition cards. A second Java application showed temporal evolution signals that were refreshed in an automated way on the screen after each shot. A third Java application provided access to the operation logbook. In addition to these tools, we used the VRVS video conferencing system (FUSION community, X-Point room) and the EFDA Messenger Service for instant messaging (Jabber client). (Author)

  20. TJ-II operation tracking from Cadarache

    International Nuclear Information System (INIS)

    The TJ-II remote participation system was designed to follow the TJ-II discharge production, even allowing the physicist in charge of operation to be in a remote location. The system has been based on both web servers and Java technology. These elements were chosen due to its open character, security properties, platform independency and technological maturity. Web pages and Java applications permit users to access experimental systems, data servers and the operation logbook. Security resources are provided by the PAPI system, a distributed authentication and authorization system. The TJ-II remote participation tools have allowed us to command and follow the stellarator operation from Cadarache. Over 1,000 digitizer channels and more than 20 diagnostic control systems were remotely available from web pages for monitoring/programming purposes. One Java application provided on-line information about the acquisition status of channels and acquisition cards. A second Java application showed temporal evolution signals that were refreshed in an automated way on the screen after each shot. A third Java application provided access to the operation logbook. In addition to these tools, we used the VRVS videoconferencing system (FUSION community, X-Point room) and the EFDA Messenger Service for instant messaging (Jabber client). (author)

  1. Corium behaviour research at CEA Cadarache. The PLINIUS prototypic corium experimental platform

    International Nuclear Information System (INIS)

    In the hypothetical case of a Light Water Reactor severe accident, the reactor core could melt and form with structure materials, and eventually concrete, a complex mixture called corium. The knowledge of corium behaviour is one of the key to increase the safety of the nuclear reactors. In the framework of severe accident studies, CEA has undertaken for many years a large RandD program devoted to this topic. At CEA-Cadarache, a specific platform, called PLINIUS, has been developed and is dedicated to studies of prototypic corium (i.e. with uranium dioxide). In this paper, the PLINIUS platform four facilities and the experiments on corium behaviour are described.(author)

  2. The ITER management advisory committee (MAC) meeting in Cadarache

    International Nuclear Information System (INIS)

    The ITER management advisory committee meeting was held on 8 March in Cadarache, France. The main topics were the ITER EDA Status Report in the period between the ITER Meeting in Yokohama (October 1990) and February 1999. In particular, the management advisory committee shares the director's concern about the uncertainties against which the whole project operates at present. They also noted that the definition of an appropriate framework for continued US involvement in on-going projects is a very important issue

  3. Meeting of the ITER CTA Project Board at Cadarache

    International Nuclear Information System (INIS)

    The meeting of the ITER CTA Project Board took place on the occasion of the N4 Meeting at Cadarache, France on 3-4 June 2002. The Project Board members and experts from Canada, EU, Japan and Russian Federation and the International Team attended the meeting. The parts informed each other of the latest developments concerning the ITER project in their respective countries, among them, offers for ITER potential site from the EU and Japan

  4. Synthetic Peptides as Structural Maquettes of Angiotensin-I Converting Enzyme Catalytic Sites

    Directory of Open Access Journals (Sweden)

    Zinovia Spyranti

    2010-01-01

    Full Text Available The rational design of synthetic peptides is proposed as an efficient strategy for the structural investigation of crucial protein domains difficult to be produced. Only after half a century since the function of ACE was first reported, was its crystal structure solved. The main obstacle to be overcome for the determination of the high resolution structure was the crystallization of the highly hydrophobic transmembrane domain. Following our previous work, synthetic peptides and Zinc(II metal ions are used to build structural maquettes of the two Zn-catalytic active sites of the ACE somatic isoform. Structural investigations of the synthetic peptides, representing the two different somatic isoform active sites, through circular dichroism and NMR experiments are reported.

  5. Taking Human and organizational factors into account in AREVA NC Cadarache dismantling projects

    International Nuclear Information System (INIS)

    AREVA NC Cadarache fabricated MOX fuel (fuel assemblies made with a mixture of uranium and plutonium oxides) and FBR fuel (fast breeder reactor) for 40 years. In 2003, commercial fuel fabrication was stopped and the plant began implementing a plan to remove the nuclear material and clean up its two production buildings and laboratory. The objective is to reduce the radiological activity in the process facilities to IAEA level 2 and transferred them back to the owner French Atomic Energy Commission (CEA). The type of facilities to be dismantled (glove boxes, tanks, etc.) and the products involved (uranium and plutonium), in the form of powder, require a high degree of manual operation. Human and Organizational Factors (HOF) therefore play a key role in the depth defense concept. (author)

  6. Technical Aspect of Shielded SIMS Installation in CEA Cadarache

    International Nuclear Information System (INIS)

    A shielded IMS 6f has been installed in the LECA facility, CEA Cadarache France. The nuclearization was performed by CAMECA Company, which sells the standard IMS 6f. Working on nuclear materials requires in depth modifications of the apparatus itself. Despite these modifications, the shielded SIMS has the same level of performance as the standard apparatus. The design of the modified apparatus is presented and the safety aspects are emphasised. The shielded SIMS should be allowed to handle irradiated samples at the end of 2001. (Author)

  7. Intensive post operation clean out (Poco) in AREVA NC Cadarache dismantling projects: economic value and sustainable development

    International Nuclear Information System (INIS)

    After fabricating plutonium fuels for 40 years (Fast Breeder Reactor, MOX), commercial operations at the AREVA NC Cadarache plant ended in 2003 and post operation clean out of its production buildings and laboratory started. AREVA objective is to achieve IAEA level 2 cleanup of the process facilities and transfer them to their owner, the French Atomic Energy Commission (CEA). There are two main opposite dismantling scenarios: very limited cleanup to optimize man power, or intensive cleanup of the facilities to optimize the quantity of long lived level waste as much as possible. Economic and environmental considerations lead the choice between these two options. (author)

  8. Report on transparency and nuclear safety 2014 - Cadarache CEA centre

    International Nuclear Information System (INIS)

    This document proposes, first, a presentation of the Cadarache CEA centre, of its activities and installations, gives a rather detailed overview of measures related to safety and to radiation protection within these activities and installations. Then it reports significant events related to safety and to radiation protection which occurred in 2014 and have been declared to the ASN. Next, it discusses the results of release measurements (liquid and gaseous effluents, radiological assessment, and chemical assessment for various installations) and the control of the chemical and radiological impact of these gaseous and liquid effluents on the environment. Finally, it addresses the issue of radioactive wastes which are stored in the different nuclear base installations of the Centre, indicates the different measures aimed at limiting the volume of these warehoused wastes and addresses their impact on health and on the environment. Nature and quantities of warehoused wastes are specified. Remarks and recommendations of the CHSCT are given

  9. Processing of radioactive effluents in Cadarache research centre

    International Nuclear Information System (INIS)

    French Atomic Energy Commission (Commissariat a l'Energie Atomique - CEA) is studying the design of a new plant for processing liquid aqueous radioactive effluents produced on Cadarache Research Centre. Effluents to be processed are low and medium activity effluents, with, for some of them, important concentrations of actinides, and which lead to the production of A and B category solid wastes, according to the French legislation, and suitable for final disposal. The objectives in terms of minimization of discharge's activity level and solid waste production have guided a selection of processes and their arrangement, in order to optimise decontamination and volume concentration factors. Seeded-ultrafiltration and vitrification with cold crucible melter are part of the selected processes for which developments are in progress in CEA for application to the characteristics of the project. (authors)

  10. Tenth ITER negotiations and related meetings [Cadarache, France, September 2005

    International Nuclear Information System (INIS)

    With the issue of ITER siting settled at the Ministerial Meeting for ITER in Moscow of 28 June 2005, detailed joint work towards all the other aspects of the agreement on the implementation of the ITER international fusion energy research project resumed as delegations from China, the European Union, Japan, the Republic of Korea, the Russian Federation and the United States of America met at the agreed site of Cadarache in the period 7-15 September 2005 for intensive working sessions culminating in the tenth official meeting of the ITER Negotiators. The major advances made were reported to the Tenth Official Meeting of the Negotiators on 12 September at which all delegations reiterated their intention to proceed together with the construction of ITER at the earliest date, and expressed their satisfaction on progress made in defining the process for designation of top ITER officials. In particular, the delegations expressed the wish to see Japan identify suitably qualified candidate(s) to be designated as Nominee Director General. In a separate meeting, the ITER delegates had an informal exchange with a delegation from India on India's interest in the possibility of participating in ITER. The Committee took a number of organizational decisions aimed at preparing for the start of construction. In particular it was agreed to establish, under the Transitional Arrangements, an ITER Joint Work Site at Cadarache. This now provides the institutional frame to start assembling international personnel and to undertake joint technical work at the ITER Site. In addition, the CEA was mandated to represent ITER, in collaboration with the International Team, in the public debate now in progress in France and in the regulatory and licensing process until the ITER Organization itself will be in a position to take the role over

  11. The giant superconducting magnet system of 10,000 tons mass for fusion experiment at Cadarache, France

    International Nuclear Information System (INIS)

    The International Thermonuclear Experimental Reactor (ITER) being built at Cadarache, France has many unique features and is one of the biggest scientific adventures in the history of science and technology. Seven partners (India, EU, US, China, Japan, Korea and Russia) have made an International Organization situated at Cadarache, France to provide direction and co-ordination for R and D and construction of this project. The R and D labs and manufacturing industries are spread in these seven partner countries. Components manufactured in these countries will be transported to Cadarache in France for assembly. Institute for Plasma Research, Bhat, Gandhinagar, Gujarat is coordinating this project activities on behalf of India. The magnet system, required for confinement and control of plasma leading to fusion reaction in ITER is one of the key systems of this project. There are 18 TF (Toroidal Field) Coils, 6 PF (Poloidal Field) coils, 6 CS (Central Solenoid) coils and 18 correction coils (CC), all of which are of superconducting type. All TF and CS coils have Nb3Sn superconductor and all PF and CC coils have NbTi superconductor. Each TF coil has height 15 m and width 9 m and 330 tons mass. The biggest PF coil has diameter 24 m and 300 tons mass. The total mass of these superconducting magnet systems is about 10000 tons. Use of Nb3Sn superconductor for manufacturing superconducting cables for successful use had not reached a matured stage earlier and this project gave a thrust for significant R and D activities worldwide and now due to this project, it is a matured and reliable technology. The jacketing and manufacturing of long cables need up to about 760 m long special infrastructure at Industry. The special building built for PF coil winding at ITER, Cadarache site is of size 250 m X 45 m. All these coils are made using cable-in-conduit conductors (CICC). These long CICCs have to carry current as high as 68 kA in case of TF coils. Due to this high current and

  12. Report on transparency and nuclear safety - Cadarache CEA centre - 2012

    International Nuclear Information System (INIS)

    A first volume proposes a presentation of the Cadarache CEA centre, of its activities and installations, gives a rather detailed overview of measures related to safety and to radiation protection within these activities and installations. It also reports significant events related to safety and to radiation protection which occurred in 2012 and have been declared to the ASN. It discusses the results of release measurements (liquid and gaseous effluents, radiological assessment, and chemical assessment for various installations) and the control of the chemical and radiological impact of these gaseous and liquid effluents on the environment. It addresses the issue of radioactive wastes which are stored in the different nuclear base installations of the Centre, indicates the different measures aimed at limiting the volume of these warehoused wastes and addresses their impact on health and on the environment. Nature and quantities of warehoused wastes are specified. The second volume concerns some specific installations (INB 32 or ATPu, and INB 54 or LPC) which belong to AREVA NC. The same topics are addressed: presentation of the facilities, arrangements regarding safety and radiation protection, significant events related to safety and radiation protection, measurements of effluents and their impact on the environment, warehoused wastes. Remarks and recommendations of the CHSCT are given

  13. Breeder reactors

    International Nuclear Information System (INIS)

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components

  14. Seismic risk and safety of nuclear installations. A look at the Cadarache Centre; Risque sismique et surete des installations nucleaires. Regards sur le Centre de Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    Verrhiest-Leblanc, G.; Chevallier, A. [Autorite de Surete Nucleaire 75 - Paris (France)

    2010-11-15

    After a brief recall of some important seismic events which occurred in the past in the south-eastern part of France, the authors indicate the nuclear installations present in this region. They outline the difference between requirements for a usual building and for basic nuclear installations. They indicate laws and regulations which are to be applied to these installations like to any hazardous industrial installation. They describe the seismic risk as it has been determined for the Cadarache area, and evoke the para-seismic design of new nuclear installations which are to be built in Cadarache and actions for a para-seismic reinforcement of existing constructions. Finally, they evoke organisational aspects (emergency plans) and the approach for a better information and transparency about the seismic risk

  15. Enforcement of control decree for radiation protection devices on the Cadarache center

    International Nuclear Information System (INIS)

    On the Cadarache center there are nearly 1800 mobile devices and 800 fixed. The periodic calibration checks are performed by type with the same sources on every installation of the Center. A commune methodology was developed by type of device. It is presented in the form of cards with a phasing of operations as a picture. (N.C.)

  16. An Information Building on Radioactivity and Nuclear Energy for the French CEA Cadarache Research Center - 13492

    International Nuclear Information System (INIS)

    The CEA Cadarache research center is one of the 10 research centers of the French Alternative Energies and Atomic Energy Commission (CEA). Distributed throughout various research platforms, it focuses on nuclear fission, nuclear fusion, new energy technologies (hydrogen, solar, biomass) and fundamental research in the field of vegetal biology. It is the most important technological research and development centers for energy in Europe. Considering the sensitive nature of nuclear activities, the questions surrounding the issue of radioactive waste, the nuclear energy and the social, economic and environmental concerns for present and future generations, the French Government asked nuclear actors to open communication and to give all the information asked by the Local Information Commission (CLI) and the public [1]. In this context, the CEA Cadarache has decided to better show and explain its expertise and experience in the area of nuclear energy and nuclear power plant design, and to make it available to stakeholders and to the public. CEA Cadarache receives each year more than 9000 visitors. To complete technical visits of the research facilities and laboratories, a scientific cultural center has been built in 2011 to inform the public on CEA Cadarache research activities and to facilitate the acceptance of nuclear energy in a way suited to the level of knowledge of the visitors. A modern interactive exhibition of 150 m2 allows visitors to find out more about energy, CEA Cadarache research programs, radioactive waste management and radiological impact on the research center activities. It also offers an auditorium for group discussions and for school groups to discover science through enjoyment. This communication center has received several thousand visitors since its opening on October 2011; the initial results of this experience are now available. It's possible to explain the design of this exhibition, to give some statistics on the number of the visitors, their

  17. Literature survey 'Impact of the Cadarache Centre activity on the environment'. Survey performed by the CRIIRAD laboratory for the Cadarache's CLI

    International Nuclear Information System (INIS)

    After a presentation of the Cadarache Centre installations and activities, this document reports a literature survey based on the documents provided by the Centre and related to dismissals and incidents since the beginning of the Centre activity, and to environment control during some critical periods. Several issues are discussed: the radioactive atmospheric effluents and their impact, the control of underground waters, the control of surface water environment (impact of liquid radioactive effluents and monitoring of run-off waters), soil and food chain monitoring. For each of these themes, control and monitoring data are analysed

  18. Continuously tritium monitoring of the pipe of liquid effluents at the Cea Cadarache; Controle en continu du tritium de la conduite des effluents liquides du CEA Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    Pira, Y

    2004-07-01

    This report is oriented toward the radiation protection of environment that is an essential component of radiation protection. It is necessary to detect any solid, liquid or gaseous abnormal release and to find its origin. The present study bears on a detection instrument in continuously to find tritium in liquid effluents of the Cea Cadarache. After having study the functioning principle of this device, an evaluation of its performances has been realised ( back noise, yield, detection limit) and to a checking in real conditions of utilization. (N.C.)

  19. 2006 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety: safety organisation, general measures, measures related to various risks, control of emergency situations, inspections, audits and second level control, measures in basic nuclear installations. It describes the measures concerning the radioprotection on this site: organisation, significant facts, and dose measurement results. It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of release measurements and of radiological and chemical assessments of the impact these releases on the environment. The report then describes measures implemented to limit the volume of stored radioactive wastes and also their impact on health and on the environment. It proposes a series of tables indicating the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation), for which the significant events occurred on the 19. of January and on the 6. of November 2006. For these installations, release measurements concern gaseous and liquid releases

  20. 2008 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety: safety organisation, general measures, measures related to various risks, control of emergency situations, inspections, audits and second level control, measures in basic nuclear installations. It describes the measures concerning the radioprotection on this site: organisation, significant facts, and dose measurement results. It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of release measurements and of radiological and chemical assessments of the impact these releases on the environment. The report then describes measures implemented to limit the volume of stored radioactive wastes and also their impact on health and on the environment. It proposes a series of tables indicating the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation) for which significant events occurred on the 18. and on the 21. of November 2008. For these installations, release measurements concern gaseous and liquid releases

  1. 2007 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety: safety organisation, general measures, measures related to various risks, control of emergency situations, inspections, audits and second level control, measures in basic nuclear installations. It describes the measures concerning the radioprotection on this site: organisation, significant facts, and dose measurement results. It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of release measurements and of radiological and chemical assessments of the impact these releases on the environment. The report then describes measures implemented to limit the volume of stored radioactive wastes and also their impact on health and on the environment. It provides a series of tables indicating the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. It reports the recommendations expressed by the CHSCT (committee on hygiene, security and working conditions) after the 2006 report. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation), for which the significant events occurred on the 13. of March and on the 25. of May 2007. For these installations, release measurements concern gaseous and liquid releases

  2. 2009 transparency and nuclear safety report. CEA Cadarache. Volumes 1 + 2

    International Nuclear Information System (INIS)

    After a general presentation of the Cadarache site and of its nuclear installations, the first volume of this report describes the various measures concerning the site safety (safety organisation, general measures, measures related to various risks, inspections, control of emergency situations, audits and second level control, measures in basic nuclear installations) and radioprotection (organisation, significant facts, dosimeter results). It describes significant events which occurred in relationship with nuclear safety and radioprotection, presents results of measurements of releases and of their impact on the environment (chemical and radiological assessment). Then after a description of measures to limit the volume of stored radioactive wastes and their impact on health and on the environment, tables indicate the nature and quantities of wastes which are stored in the different basic nuclear installations of Cadarache. The second volume proposes the same information for two specific nuclear installations belonging to Areva and located in Cadarache, the INB 32 and 54 (INB stands for basic nuclear installation) for which a significant event occurred on the 6 October 2009. For these installations, release measurements concern gaseous and liquid releases

  3. Influence of high-permeability discs in an axisymmetric model of the Cadarache dynamo experiment

    Science.gov (United States)

    Giesecke, A.; Nore, C.; Stefani, F.; Gerbeth, G.; Léorat, J.; Herreman, W.; Luddens, F.; Guermond, J.-L.

    2012-05-01

    Numerical simulations of the kinematic induction equation are performed on a model configuration of the Cadarache von-Kármán-sodium dynamo experiment. The effect of a localized axisymmetric distribution of relative permeability μr that represents soft iron material within the conducting fluid flow is investigated. The critical magnetic Reynolds number Rmc for dynamo action of the first non-axisymmetric mode roughly scales like Rmcμr - Rmc∞∝μ-1/2r, i.e. the threshold decreases as μr increases. This scaling law suggests a skin effect mechanism in the soft iron discs. More important with regard to the Cadarache dynamo experiment, we observe a purely toroidal axisymmetric mode localized in the high-permeability discs which becomes dominant for large μr. In this limit, the toroidal mode is close to the onset of dynamo action with a (negative) growth rate that is rather independent of the magnetic Reynolds number. We qualitatively explain this effect by paramagnetic pumping at the fluid/disc interface and propose a simplified model that quantitatively reproduces numerical results. The crucial role of the high-permeability discs in the mode selection in the Cadarache dynamo experiment cannot be inferred from computations using idealized pseudo-vacuum boundary conditions (H × n = 0).

  4. [Johan Engström. Revals Befästningsdhistoria sedd utifrån en maquett från 1682. Statens försvarshistoriska museer 2011. Skrift nr 17. ] / Ragnar Nurk, Robert Treufeld

    Index Scriptorium Estoniae

    Nurk, Ragnar

    2014-01-01

    Arvustus: Johan Engström. Revals Befästningsdhistoria sedd utifrån en maquett från 1682. Statens försvarshistoriska museer 2011. Skrift nr 17. 2011. Rootsi-aegsestest kindlustustest. Hilisema perioodi kindlustustest

  5. Teaching Sodium Fast Reactors in CEA-INSTN

    International Nuclear Information System (INIS)

    Conclusion: Education and Training: - a key element for the future of the development of Sodium Fast Reactors, and more particularly ASTRID project. - a tool to create a new generation of skilled nuclear engineers in the field. - a unique mean to share basic knowledge, operational feedback, safety approaches. The two entities aimed to deliver Training sessions, i.e. Sodium School in Cadarache, and INSTN-Cadarache, are ready: - to conceive and propose tailored sessions, - to collaborate with other foreign Education and Training Entities

  6. Results of a large scale neutron spectrometry and dosimetry comparison exercise at the Cadarache moderator assembly

    International Nuclear Information System (INIS)

    Eurados Working Group 7 recently organised a large-scale comparison of neutron spectrometry and dosimetry measurements at the IPSN/SDOS laboratory of the CEA Cadarache Research Centre in France. A large number of participants took part with a range of instruments including spectrometers, tissue-equivalent proportional counters, personal dosemeters, and survey instruments. The neutron field used for the exercise was a primarily low energy neutron spectrum similar to those which have been measured recently around nuclear facilities. This paper presents the results of the measurements and attempts to draw conclusions about the accuracy attainable with the various devices, their advantages and drawbacks, and potential problems. (author)

  7. Influence of high permeability disks in an axisymmetric model of the Cadarache dynamo experiment

    OpenAIRE

    A. Giesecke; Nore, C.; Stefani, F.; Gerbeth, G.; Léorat, J.; Herreman, W; Luddens, F.; Guermond, J.-L

    2011-01-01

    Numerical simulations of the kinematic induction equation are performed on a model configuration of the Cadarache von-K\\'arm\\'an-Sodium dynamo experiment. The effect of a localized axisymmetric distribution of relative permeability {\\mu} that represents soft iron material within the conducting fluid flow is investigated. The critical magnetic Reynolds number Rm^c for dynamo action of the first non-axisymmetric mode roughly scales like Rm^c({\\mu})-Rm^c({\\mu}->infinity) ~ {\\mu}^(-1/2) i.e. the ...

  8. French fission products experiments performed in Cadarache and Valduc. Results comparison

    International Nuclear Information System (INIS)

    Cofunded by Cogema, two complementary experimental programmes on burn up credit (BUC) related to fission products (FPs) are performed by CEA and IRSN at Cadarache and Valduc. After shortly recalling the main characteristics of each experiment, a first comparison of some results is presented, especially the energy range in which most part of cross section absorption are qualified. Both experiments exhibit great quality and accurate results, giving a high degree of confidence to the whole experimental French process of qualification devoted to BUC. (author)

  9. Influence of high permeability disks in an axisymmetric model of the Cadarache dynamo experiment

    CERN Document Server

    Giesecke, A; Stefani, F; Gerbeth, G; Léorat, J; Herreman, W; Luddens, F; Guermond, J -L

    2011-01-01

    Numerical simulations of the kinematic induction equation are performed on a model configuration of the Cadarache von-K\\'arm\\'an-Sodium dynamo experiment. The effect of a localized axisymmetric distribution of relative permeability {\\mu} that represents soft iron material within the conducting fluid flow is investigated. The critical magnetic Reynolds number Rm^c for dynamo action of the first non-axisymmetric mode roughly scales like Rm^c({\\mu})-Rm^c({\\mu}->infinity) ~ {\\mu}^(-1/2) i.e. the threshold decreases as {\\mu} increases. This scaling law suggests a skin effect mechanism in the soft iron disks. More important with regard to the Cadarache dynamo experiment, we observe a purely toroidal axisymmetric mode localized in the high permeability disks which becomes dominant for large {\\mu}. In this limit, the toroidal mode is close to the onset of dynamo action with a (negative) growth-rate that is rather independent of the magnetic Reynolds number. We qualitatively explain this effect by paramagnetic pumping...

  10. Cadarache, January 17, 2012. A nuclear crisis exercise with a seismic dimension

    International Nuclear Information System (INIS)

    As France is a region with a moderate seismicity, the new French seismic zoning indicates that the Cadarache CEA centre and the surrounding communities are located in a region exhibiting a seismicity which is greater than the metropolitan average. Therefore, a project of organizing a nuclear crisis exercise emerged with taking this seismic risk into account. This exercise took place on the 17 January 2012. This article presents its objectives which addressed the ability of public authorities and of the operator to manage a nuclear accident caused by an earthquake and its consequences in terms of civilian security and nuclear safety. The article also briefly presents the Cadarache centre within its seismic environment, and how the seismic risk had been taken into account when designing and dimensioning some of its installations. It also evokes the ASN prescriptions made after the decennial safety inspections. Several actors, institutions and organizations have been involved in the exercise and in its preparation, and this preparation had some consequences on the awareness of local authorities on the seismic risk. The adopted scenario is briefly presented as well as the performance of simulations. The main lessons learned are briefly discussed

  11. Web software for the control and management of radiation protection devices in the Cadarache site

    International Nuclear Information System (INIS)

    This series of slides presents how to use a new software dedicated to the management of the periodical controls that have to be performed on the equipment involved in radiation protection. This software is ready to be dispatched on the CEA site of Cadarache. This software gives information on: the device to be controlled, the controls that have to be performed, the procedures to follow to make the test, the equipment necessary for the test particularly the need for radioactive sources, the maintenance of the device, the previous measurements and in the end the device's conformity. An evaluation of the conformity of all the devices present in a building or an area or of a particular type can be easily obtained. (A.C.)

  12. Evolution of ITER Tritium Confinement strategy and adaptation to Cadarache site conditions and French regulatory requirements

    International Nuclear Information System (INIS)

    The ITER Nuclear Buildings include the Tokamak, Tritium and Diagnostic Buildings (Tokamak Complex) and the Hot Cell and Low Level Radioactive Waste Buildings (Hot Cell Complex). The Tritium Confinement strategy of the Nuclear Buildings comprises key features of the atmosphere and vent detritiation systems and the heating, ventilation and air conditioning systems. The designs developed during the ITER EDA (engineering design activities) for these systems need to be adapted to the specific conditions of the Cadarache site and modified to conform with the regulatory requirements applicable to installations nucleaires de base (INB) - basic nuclear installations - in France. The highest priority for such adaptation has been identified as the Tritium Confinement of the Tokamak Complex and the progress in development of a robust, coherent design concept compliant with French practice is described in the paper

  13. NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

    International Nuclear Information System (INIS)

    To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g-1. Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the

  14. Neutron measurements in MASURCA reactor: preliminary experience

    International Nuclear Information System (INIS)

    In the context of the studies on ADS, an experimental program is in progress in collaboration with the CEA team of the small nuclear reactor MASURCA at Cadarache. We report the first test of neutron flux measurement and the status of the program. (authors)

  15. KROTOS FCI experimental programme at CEA Cadarache: new features and status

    Energy Technology Data Exchange (ETDEWEB)

    Bonnet, J.M.; Bullado, Y.; Journeau, C.; Fouquart, P.; Piluso, P.; Sergeant, C.; Magallon, D. [CEA-Cadarache, DTN/STRI/LMA, Bat 708, BP1, 13108 Saint Paul lez Durance cedex (France)

    2005-07-01

    Full text of publication follows: KROTOS facility has been operated by the European Commission at JRC-Ispra for many years until 1999 as part of the FARO/KROTOS programme. The programme had to be stopped at JRC due to new EC priorities, and an agreement was concluded with CEA to transfer the KROTOS facility and know-how to Cadarache to continue investigating the pending FCI issues. By this agreement, CEA became owner of the facility. The main objective of the KROTOS programme at CEA Cadarache is understanding the role of melt properties in steam explosion propagation and energetics and, in particular, steam explosion behaviour of prototypical corium melts. Possible influence of physicochemical processes on explosiveness will also be explored. Actually, analysis of alumina debris produced at Ispra have shown that formation of metastable phases and chemical reaction with water at high temperature may play a significant role in enhancing heat transfer to water in the explosion phase. In order to reach the objective, steam explosion experiments are performed in well characterised conditions for a large spectrum of conditions and melt compositions of interest for both in- and ex-vessel situations. A trigger is applied as a rule. Advanced technology and instrumentation is used to reduce uncertainties on initial conditions and characterise the various phases of an explosion, with emphasis on high energy X-ray cinematography to qualify pre-mixing. This advanced instrumentation will enable the measurement of detailed variables to consolidate the qualification of the FCI codes. The use of high energy X-rays made it necessary to construct a new building to house the facility and its components. The facility should newly become operative early 2005. Main improvements with respect to Ispra concern melt delivery, hydrogen measurement and X-ray imaging. Preliminary studies have demonstrated that identification of the pre-mixture-water interface and coherent melt jet core, and

  16. Flow modelling in fractured aquifers, development of multi-continua model (direct and inverse problems) and application to the CEA/Cadarache site

    International Nuclear Information System (INIS)

    This research thesis concerns the modelling of aquifer flows under the CEA/Cadarache site. The author reports the implementation of a numerical simulation tool adapted to large scale flows in fractured media, and its application to the Cadarache nuclear site. After a description of the site geological and hydrogeological characteristics, the author presents the conceptual model on which the modelling is based, presents the inverse model which allows a better definition of parameters, reports the validation of the inverse approach by means of synthetic and semi-synthetic cases. Then, he reports experiments and simulation of the Cadarache site

  17. LWR Fuel Gas Characterization at CEA Cadarache LECA-STAR Hot Laboratory

    International Nuclear Information System (INIS)

    The aim to improve LWR fuel behaviour led CEA to improve its PIE capacities in term of test devices and characterization techniques in the shielded hot cells of the LECA-STAR facility, located in Cadarache. A presentation of these capacities is made, focusing on gas characterization : - annealing test devices used for transient simulations. These devices are test facilities allowing to simulate a very large range of conditions (temperature, atmosphere, pressure etc); - characterization techniques : optical microscopy, for microstructure characterizations on polished samples; scanning electron microscopy (SEM) used both on polished samples and on fractographs; Electron Probe Micro-Analyzer (EPMA) for quantitative elementary analysis and elementary mappings; Secondary Ion Mass Spectrometer (SIMS), for isotopic analysis, mapping depth profiles and gas measurements. It shows an overview of their main applications for the study of fuel behaviour under nominal operating conditions but also under simulated accidental conditions and storage conditions. Two detailed examples of the use of these techniques working together on the same samples, are then presented, focusing on fission gas characterizations. The first one is a 72 GW.d.t-1 high burnup UO2 PWR fuel for which detailed characterizations have been performed before and after a LOCA type condition annealing test. The second one is a 35 GW.d.t-1 UO2 PWR fuel for which these characterizations have been done before and after a ramp test at 100 W.cm-1.min-1 up to a 90 s maximum power hold time at 520 W.cm-1. In both cases, the complementarities of these various techniques show the interest of such detailed characterizations. (author)

  18. The jules Horowitz reactor: experimental possibilities

    International Nuclear Information System (INIS)

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

  19. Report on transparency and nuclear safety, Cadarache 2011, Volume 1 + Volume 2 INB 32 and 54 Areva NC

    International Nuclear Information System (INIS)

    For respectively different sites all situated in Cadarache, the CEA Centre on the one hand, and the AREVA nuclear base installation INB 32 and 54 on the other hand, these volumes propose a presentation of their activities and installations, a description of measures regarding safety and radiation protection, a description of significant events regarding nuclear safety and radiation protection, a presentation and discussion of results of measurements of releases and of their impact on the environment, and a description of the different radioactive wastes stored on the concerned site

  20. Rapport d'étude : La perception des risques majeurs par les riverains du CEA de Cadarache

    OpenAIRE

    Lopez, Alexia; Régner, Isabelle; Schleyer-Lindenmann, Alexandra

    2015-01-01

    Ce rapport d'étude vient répondre aux interrogations de son commanditaire, la CLI (Commission Local d'Information) du CEA (Commissariat à l'énergie Atomique) de Cadarache. Il s'organise en trois partie : une première partie établit un état des lieux des connaissances des risques majeurs par les habitants. Les risques majeurs officiellement recensés sont : le risque feu de forêt, le risque inondation, le risque sismique, le risque de mouvement de terrain, le risque de rupture de barrage, et le...

  1. Dossier: research reactors

    International Nuclear Information System (INIS)

    Research reactors are used at the CEA (the French atomic energy commission) since many years. Their number has been reduced but they remain unique tools that CEA valorize continuously. The results of the programs involving such reactors are of prime importance for the operation of Electricite de France (EdF) park of existing power plants but also for the design of future nuclear power plants and future research reactors. This dossier presents three examples of research reactors in use at the CEA: Osiris and Orphee (CEA-Saclay), devoted to nuclear energy and fundamental research, respectively, and the critical mockups Eole, Minerve and Masurca (CEA-Cadarache) devoted to nuclear data libraries and neutronic calculation. (J.S.)

  2. Étude, réalisation et applications d'une maquette polyvalente de développement ARM7 enseignement-recherche

    Directory of Open Access Journals (Sweden)

    Pradel Gilbert

    2011-06-01

    Full Text Available Cet article présente une expérience de transfert bidirectionnel recherche-enseignement en informatique industrielle. L'expérience accumulée dans le cadre du projet ANR RobAutiSTIC [1] et la confrontation des besoins spécifiques en matière de maquettes d'expérimentation en recherche et dans l'apprentissage de l'informatique industrielle au niveau licence/maîtrise ont guidé nos choix dans la réalisation d'une carte polyvalente autour d'un processeur à coeur ARM7 [2]. Cette carte a été conçue pour être compatible avec la base robotique du projet RobAutiSTIC et servir de support à la mise en place d'un enseignement en architecture des systèmes embarqués et enfouis et en systèmes multitâche temps réel. Cet enseignement s'insère dans la préparation à l'agrégation de génie électrique de l'ENS de Cachan et dans la formation M2P en Génie des Systèmes Informatiques de l'université Paris XI, centre d'Orsay. Le transfert initial recherche vers enseignement s'est accompagné d'un retour de l'enseignement vers la recherche : certains objectifs pédagogiques ont conduit à mettre en oeuvre de nouvelles fonctionnalités et à des simplifications qui ont bénéficié au robot du projet RobAutiSTIC.

  3. Decommissioning of the Rapsodie fast reactor: developing a strategy

    International Nuclear Information System (INIS)

    The RAPSODIE experimental fast neutron reactor at Cadarache (France) was operated from 1962 to 1982. The initial decommissioning operations began immediately, reaching IAEA stage 2 in 1994. Since then, the facility has been maintained under surveillance pending final dismantling scheduled to begin in 2020. New studies are now in progress to accelerate the dismantling process. The present status of the reactor block is described, the advantages and drawbacks of early dismantling are considered, and various dismantling scenarios are discussed. (author)

  4. Prediction of Long Term Corrosion Behaviour in Nuclear Waste Systems. Proceedings of an International Workshop, Cadarache, France, 2002

    International Nuclear Information System (INIS)

    This 36. book of the European Federation of Corrosion series contains the proceedings of the International Workshop on 'Prediction of Long Term Corrosion Behaviour in Nuclear Waste Systems' that took place in Cadarache, France in November 2001. The EFC Working Party on Nuclear Corrosion, which sponsored this Workshop, had already considered in 1991 that the long term integrity of Radioactive Waste storage was a crucial problem in the development of nuclear energy for electricity production. At that time, it was realized that public opinion on Nuclear Energy, was concerned not only with the risk of a nuclear accidents but also with the safety of long term storage of high level nuclear wastes. The nuclear community and the public demanded that any technical barrier for preventing long term radionuclides from entering the biosphere should be effective and guaranteed for 10,000 or more. The integrity of the waste containers, particularly their resistance to possible internal and external corrosion, is the most important aspect of this requirement. For the last ten years, the importance of programmes on long term corrosion on nuclear waste canisters has become of major significance to the international community. The Cadarache Workshop included reports on these programmes from eight countries. These approached the long term corrosion tests, but also by the development of the necessary deterministic and empirical models for extrapolating the behaviour of the selected materials. The difficulties involved in developing such models of corrosion behaviour and in demonstrating their validity has been underlined many times. The Proceedings contains in all an Introduction by Professor G. Beranger and 33 papers grouped in four parts as follows: 1 - R and D corrosion programmes (5 papers); 2 - Laboratory and in situ testing (11 papers); 3 - Historical and archaeological analogues (5 papers); and 4 - Fundamental issues, models and prediction

  5. Reactors

    International Nuclear Information System (INIS)

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

  6. Jules Horowitz reactor, basic design

    International Nuclear Information System (INIS)

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: represent a significant step in term of performances and experimental capabilities; be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements; reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (author)

  7. Jules Horowitz Reactor, basic design

    International Nuclear Information System (INIS)

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: - represent a significant step in term of performances and experimental capabilities, - be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements, - reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (authors)

  8. Teaching Sodium Fast Reactor Technology and Operation for the Present and Future Generations of SFR Users

    OpenAIRE

    Christian, Latge; Rodriguez, Gilles; Baque, Francois; Leclerc, Arnaud; Martin, Laurent; Vray, Bernard; Romanetti, Pascale

    2011-01-01

    This paper provides a description of the education and training activities related to sodium fast reactors, carried out respectively in the French Sodium and Liquid Metal School (ESML) created in 1975 and located in France (at the CEA Cadarache Research Centre), in the Fast Reactor Operation and Safety School (FROSS) created in 2005 at the Phenix plant, and in the Institut National des Sciences et Techniques Nucle'aires (INSTN). It presents their recent developments and the current collaborat...

  9. Decommissioning of fast reactors after sodium draining

    International Nuclear Information System (INIS)

    Acknowledging the importance of passing on knowledge and experience, as well mentoring the next generation of scientists and engineers, and in response to expressed needs by Member States, the IAEA has undertaken concrete steps towards the implementation of a fast reactor data retrieval and knowledge preservation initiative. Decommissioning of fast reactors and other sodium bearing facilities is a domain in which considerable experience has been accumulated. Within the framework and drawing on the wide expertise of the Technical Working Group on Fast Reactors (TWG-FR), the IAEA has initiated activities aiming at preserving the feedback (lessons learned) from this experience and condensing those to technical recommendations on fast reactor design features that would ease their decommissioning. Following a recommendation by the TWG-FR, the IAEA had convened a topical Technical Meeting (TM) on 'Operational and Decommissioning Experience with Fast Reactors', hosted by CEA, Centre d'Etudes de Cadarache, France, from 11 to 15 March 2002 (IAEA-TECDOC- 1405). The participants in that TM exchanged detailed technical information on fast reactor operation and decommissioning experience with various sodium cooled fast reactors, and, in particular, reviewed the status of the various decommissioning programmes. The TM concluded that the decommissioning of fast reactors to reach safe enclosure presented no major difficulties, and that this had been accomplished mainly through judicious adaptation of processes and procedures implemented during the reactor operation phase, and the development of safe sodium waste treatment processes. However, the TM also concluded that, on the path to achieving total dismantling, challenges remain with regard to the decommissioning of components after sodium draining, and suggested that a follow-on TM be convened, that would provide a forum for in-depth scientific and technical exchange on this topic. This publication constitutes the Proceedings of

  10. Reacteur Jules Horowitz (RJH): A new material testing reactor. Status report

    International Nuclear Information System (INIS)

    The 'REACTEUR JULES HOROWITZ' (RJH) is a new research reactor dedicated to material and nuclear fuel testing. This reactor, which will be erected in the CEA Cadarache nuclear research Center is now at a feasibility study stage. At the beginning of the next century, at a time when most of existing material testing reactors will have to be shutdown or will be at the end of their lifetime, le RJH will offer outstanding neutron flux levels (twice those of existing french reactors). This paper deals with the following topics - functional specifications of the project, - safety approach, - design and construction codes, - alternative designs under consideration at the feasibility stage. (author)

  11. COMAC. Nuclear data covariance matrices library for reactor applications

    International Nuclear Information System (INIS)

    Uncertainty quantification for new reactor concepts such as the ASTRID project (sodium-cooled FBR studies at CEA) requires a precise evaluation of nuclear data covariances. Since 2005, the SPRC at CEA-Cadarache has developed a covariance matrices library, called COMAC (COvariance MAtrices from Cadarache, version 0.1), for the most important nuclear data: cross sections, multiplicities (prompt and delayed neutrons), prompt fission neutron spectrum (PFNS), fission yields. This paper presents the methodology used to produce those covariance matrices, illustrated with representative cases (23Na, 238U, 239Pu, 240Pu). We have used the CONRAD evaluation code to analyse simultaneously differential and specific integral experiments and to generate a coherent and realistic set of covariances. The quality of those covariances are illustrated with uncertainty propagation calculations on key parameters for ASTRID. (author)

  12. Use of research reactors for training and teaching nuclear sciences

    International Nuclear Information System (INIS)

    Training activities on reactors are organized by Cea on 2 specially dedicated reactors Ulysse (Saclay) and Siloette (Grenoble) and 2 research reactors Minerve (Cadarache) and Azur (Cadarache, facility managed by Technicatome). About 4000 students have been trained on Ulysse since its commissioning more than 40 years ago. The concept that led to the design of Ulysse was to build a true reactor dedicated to teaching and training activities and that was able to operate with great flexibility and under high conditions of safety, this reactor is inspired from the Argonaut-type reactor. The main specificities of Ulysse are: a nominal power of 100 kW, a maximal thermal neutron flux of 1.4 1012 n.cm-2.s-1, a 90 % enriched fuel, a graphite reflector, the use of water as coolant and moderator, and 6 cadmium plates as control rods. Ulysse allows students to get practical experience on a large range of topics: approach to criticality, effect of the starting neutron source, calibration of control rods, distribution of the neutron flux in the thermal column, temperature coefficient, radiation detectors, neutron activation analysis, and radioprotection. (A.C.)

  13. The CEA Cadarache site. Additional safety assessment with respect to the accident which occurred in the Fukushima-Daiichi nuclear power station

    International Nuclear Information System (INIS)

    After a presentation of some characteristics of the CEA Cadarache site (internal and external industrial environment, crisis management organization at the CEA level and at the local level), this document reports the identification of structures and equipment concerned by crisis management (site support functions, critical structures and equipment concerned by additional safety assessments). Then, it addresses the different risks: earthquake (sizing of critical structures and equipment, margin assessment), flooding (possible origins, alarm measures), other extreme natural events (extreme meteorological conditions, extreme earthquake with induced flooding, forest fire), and loss of electric supplies and of cooling systems. The last parts address the organization of accident management in situation typically related to additional safety management), and subcontracting conditions and practices

  14. Test programme for the nuclear fuel of the Jules Horowitz reactor starts

    International Nuclear Information System (INIS)

    In Cadarache in the south of France, an advanced material test reactor is currently being built. The Jules Horowitz reactor is an initiative of the French Atomic Energy Commission ((CEA), supported by the European Commission and research centres in the Czech Republic, Spain, Finland and Belgium, plus a number of large energy companies. On the request of CEA, SCK-CEN started characterising the nuclear fuel for this reactor, in its own BR2. The article describes SCK-CEN's programme on the characterisation of new nuclear fuel elements.

  15. Radioactive waste storage at the CEA Centre in Cadarache; transfer from the -storage yard- to the CEDRA facility: an illustration of changes in the safety requirements

    International Nuclear Information System (INIS)

    A lot of changes have been made to the nuclear waste storage and disposal strategy over the last few years thanks to technological progress and changes in the requirements applicable in the sector. In this article, the authors present a concrete example of the actions undertaken, at the request of the French Nuclear Safety Authority (ASN), by the CEA Centre at Cadarache, regarding the Cadarache Centre waste storage facility (BNI 56), a facility which was originally intended as a permanent repository for solid waste. Commissioned in 1969, BNI 56 is made up of various trenches and pits which no longer meet current ASN requirements. To clean up these trenches, a legacy from the past, now requires considerable resources and involves many problems, especially due to uncertainty surrounding the nature, activity levels, packaging and volume of this waste. ASN ensures that the CEA will fully assume its overriding responsibility in the matter and implement safe, rigorous and transparent management solutions for all the waste. The BNI 56 trenches are an ongoing and concrete example of the difficulties inherent in recovering waste from storage and dismantling former facilities the design of which failed to take such aspects into account. This waste is now being repackaged so that it can be stored at a new basic nuclear installation called CEDRA (packaging and storage of radioactive waste) which complies with the safety requirements currently in force. This facility, commissioned in 2006, is intended for the storage of Type B solid radioactive waste for a period of 50 years pending construction of a deep geological repository or long-term storage facility. (authors)

  16. CONTRIBUTION OF PROTOTYPIC MATERIAL TESTS ON THE PLINIUS PLATFORM TO THE STUDY OF NUCLEAR REACTOR SEVERE ACCIDENTS

    OpenAIRE

    JOURNEAU Christophe

    2008-01-01

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium, ...). ...

  17. Report on transparency and nuclear safety 2013 - Cadarache CEA centre - Tome 1. INB 32 (ATPu) et 54 (LCP) AREVA NC - Tome 2

    International Nuclear Information System (INIS)

    A first volume proposes a presentation of the Cadarache CEA centre, of its activities and installations, gives a rather detailed overview of measures related to safety and to radiation protection within these activities and installations. It also reports significant events related to safety and to radiation protection which occurred in 2013 and have been declared to the ASN. It discusses the results of release measurements (liquid and gaseous effluents, radiological assessment, and chemical assessment for various installations) and the control of the chemical and radiological impact of these gaseous and liquid effluents on the environment. It addresses the issue of radioactive wastes which are stored in the different nuclear base installations of the Centre, indicates the different measures aimed at limiting the volume of these warehoused wastes and addresses their impact on health and on the environment. Nature and quantities of warehoused wastes are specified. The second volume concerns some specific installations (INB 32 or ATPu, and INB 54 or LPC) which belong to AREVA NC. The same topics are addressed: presentation of the facilities, arrangements regarding safety and radiation protection, significant events related to safety and radiation protection, measurements of effluents and their impact on the environment, warehoused wastes. Remarks and recommendations of the CHSCT are given

  18. The Jules Horowitz reactor: an European material testing reactor main design options

    International Nuclear Information System (INIS)

    CEA has launched an initiative for the construction in Cadarache of a second generation MTR - Jules Horowitz Reactor JHR - To be operated as an international user facility addressing major industrial and societal needs for several decades ahead. - Providing effective experimental capabilities to study material and fuel under irradiation. - Supporting safety, reliability and improved economy of commercial reactors. - Assessing future systems required for sustainable power supply and/or waste management, including future fission or fusion systems and hybrid systems. - Helping national and international decision processes on for instance nuclear safety issues or matters related to the development of nuclear energy. - Supporting education and training for future nuclear scientists and engineers and inducing structuring effects in the international nuclear community. (author)

  19. The reactor Cabri

    International Nuclear Information System (INIS)

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m3/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  20. Corrigendum to small power and heat generation systems on the basis of propulsion and innovative reactor technologies (IAEA-TECDOC-1172)

    International Nuclear Information System (INIS)

    Full text: - Footnote 1 on page 4 should read: ''1The results of the feasibility studies were presented by BARC at the meeting, but no paper was provided for publication in these proceedings.'' - On page 5, footnote 2 should be added to the following paragraph: ''CEA and TECHNICATOM (France) have about 200 reactor-years of experience from propulsion and small experimental reactors. Special design features for a future small reactor are formulated as follows2:'' (''2 At the meeting, these features were presented by CEA, Cadarache, but no paper was provided for publication in these proceedings.'')

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

    International Nuclear Information System (INIS)

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

  2. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  3. Toward a Competitive Process Intensification: A New Generation of Heat Exchanger-Reactors Vers une intensification des procédés économiquement viable : une nouvelle génération de réacteur-échangeur de chaleur

    Directory of Open Access Journals (Sweden)

    Tochon P.

    2010-10-01

    Full Text Available Process Intensification (PI in chemical production is a major concern of chemical manufacturers. Among the numerous options to intensify a process, the transposition from a batch reactor to a continuous plug flow reactor is a good alternative when the selectivity and the thermal exchange are an issue. In this context, the RAPIC R&D project aims to develop an innovative low-cost component (in the 10 kg/h range. This project deals with the design from the local to the global scale and with testing, from elementary mock-ups to pilot scale. The present paper gives a detailed description of this research project and presents the main results on specification and definition of the reaction channel and the first simple mock-ups. L’intensification des procédés (IP constitue une des préoccupations majeures de l’industrie chimique. Parmi les nombreuses options possibles pour intensifier un procédé, lorsque le sujet concerne la sélectivité et l’échange thermique, la transposition d’un réacteur discontinu à un réacteur tubulaire continu paraît une bonne alternative. Dans ce contexte, le projet de R&D RAPIC consiste à développer un composant innovant à bas coût (environ 10 kg/heure. Ce projet porte sur la phase de conception, du niveau local au niveau global, et sur la phase de tests, depuis les maquettes élémentaires jusqu’aux essais pilote. Cet article propose une description détaillée de ce projet de recherche et présente les principaux résultats obtenus sur la définition et la spécification du canal de réaction, ainsi que les premières maquettes simples.

  4. MTR fuel plate qualification in OSIRIS reactor

    International Nuclear Information System (INIS)

    Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

  5. Neutronic study regarding transmutation fuel research at Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    In order to estimate the possibilities for transmutation experiments at the Jules Horowitz Reactor several ideas for neutronic and fuel behaviour studies are investigated at CEA Cadarache. Naturally an exact replication of the burning of minor actinides in fast reactors, as expected in most transmutation scenarios, is impossible, but some key transmutation parameters can be investigated in a MTR neutron spectrum. In this paper a parametric study regarding fuel damage by He and fission products in AmUO2 is presented. By varying flux level, uranium enrichment and americium content of the sample in the JHR reflector a He production to fission ratio comparable to reference samples in the core of a SFR can be achieved. The calculations were done with the depletion code DARWIN2.2 using JEF2.2 data and spectra from a TRIPOLI model of JHR and an ERANOS model for the SFR respectively. (author)

  6. Fast breeder reactor research

    International Nuclear Information System (INIS)

    reactors of the future, the body of research aimed at developing liquid metal cooled fast reactors, national plans for work in 1976 on developing fast reactors - these were some of the topics discussed in connection with the national programmes. The development of power reactors involves a wide range of problems in the fields of nuclear and reactor physics, the thermophysics, chemistry, physics and technology of the cooling system, structural materials and nuclear fuel, the fabrication of reliable fuel elements and operating equipment, reactor monitoring and control, spent fuel reprocessing, the economics of constructing fast power reactors, nuclear safety, etc. The IWGFR, as at previous meetings, therefore paid great attention to the matter of holding international specialists' meetings. The working group recommended that the IAEA should organize the following IWGFR meetings in 1976: (1) In-Service Inspection and Monitoring (Bensberg, FRG, March 1976). (2) Cavitation in Sodium and Studies of Analogy with Water as Compared to Sodium (Cadarache, France, April 1976). (3) High Temperature Structural Design Technology (United States, May 1976) (4) Aerosol Formation, Vapour Deposits and Sodium Vapour Trapping (France, September-December 1976). The Group welcomed the IAEA's proposal to hold specialists' meetings on 'Fast Reactor Instrumentation' and 'Fuel Reprocessing and Recycling Techniques' within the framework of the Agency's programme of working groups in 1976. After discussing questions of co-ordinating and organizing international conferences on fast reactors, the IWGFR agreed to send representatives to the joint meeting of the American Nuclear Society and the American Institute of Metallurgical Engineers on 'Liquid Metal Technology', to be held at Champion, Pennsylvania, U.S.A. from 3-6 May 1976, and recommended that the IAEA should organize an international symposium on the 'Design, Construction and Operating Experience of Demonstration Fast Power Reactors' at Bologna

  7. Quantitative analysis of gases in fuels. Applications to PWR type reactors fuels

    International Nuclear Information System (INIS)

    The different methods used in Saclay and Cadarache to determine the quantity of gases which are present in fuels and fuel cans of PWR type reactors are described. These gases are fission gases (Xe, Kr), pollutant gases (hydrocarbons, N2, O2, H2O, CO, CO2), filling gases (He) or hydrides. A description of the equipment and the operation mode used are given. The obtained results on uranium oxides and mixed oxide fuels are compared with the measures of gases released in the whole rod. (O.M.)

  8. Graphite reactor physics

    International Nuclear Information System (INIS)

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm2, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors)

  9. Reactor Safety Research Programs Quarterly Report January - March 1980

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  10. REX 2000 core : a new material testing reactor project

    International Nuclear Information System (INIS)

    REX 2000 is a new research reactor project entirely dedicated to technological irradiations, which should be located on the CEA site of CADARACHE. It will be aimed at satisfying the future needs for the validation of new concepts of nuclear materials and fuels, and will take over and replace the present experimental reactors, which are 30 to 40 years old. The fundamental studies started by the CEA in 1993, on future irradiation needs expected in 2005, lead to the design of a reactor which will essentially meet the needs of PWRs, without forgetting the other fields such as FBRs, fusion... The current reactor project is based on a light water open pool concept, with a thermal power of 100 MW, in about 150 l, and characterized by an in-core-central hole. It reaches neutronic flux levels twice those of present French reactor fluxes. It allows many irradiations in the central loop under high fast neutron flux, in order to accelerate the aging of materials and analyze their behaviour. It also enables the achievement of power transient tests under high thermal neutron flux gradients. These performances are obtained with high forced flow rates and upward flow in the core, in order to preserve the operating flexibility of the reactor. This leads to the design of a specific assembly design. (author)

  11. CABRI Reactor: The fast neutron Hodoscope Calibration curves calculation with MORET

    Science.gov (United States)

    Bernard, Franck; Chevalier, Vincent; Venanzi, Damiano

    2014-06-01

    This poster presents the Hodoscope calibration curves calculation with 3D Monte Carlo code MORET. The fast neutron hodoscope is a facility of the CABRI research reactor at Cadarache (FRANCE). This hodoscope is designed to measure the fuel motion during a RIA in a pressurized water reactor. The fuel motion is measured by counting fast fission neutrons emerging from the test fuel placed in an experimental loop functioning like a Pressurized Water Reactor (T=300°C and P=155 bar), at the center of the CABRI core. The detection system of the hodoscope measures a signal which is a function of the fuel motion. The calibration curves allow then to convert the signal in a fuel mass. In order to calculate these curves, we have developed a method based on a Monte Carlo calculation code.

  12. Vibrations of the elements of a core of fast neutrons reactor to the excitation due to sodium flow

    International Nuclear Information System (INIS)

    In a fast neutron reactor, the fluid reaches the highest speeds in the core. Very turbulent, the resulting excitation incites the core elements to vibrate. A knowledge of the relative amplitudes and relative lengthening attained is necessary for the evaluation of the damage which could ultimately be produced. These vibration studies are carried out at the Cadarache Nuclear Center of the totality of the assemblies and on the constituents of these assemblies. Concurrently, an analysis of the fluctuating pressures in the fluid is carried out. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-21

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

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

    Science.gov (United States)

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

    2015-04-01

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

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

    International Nuclear Information System (INIS)

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

  16. The RES Reactor. A test reactor for the French naval propulsion

    International Nuclear Information System (INIS)

    In the Cadarache nuclear research centre the French Atomic Energy Commission (CEA) operates, with the support of TECHNICATOME as nuclear operator, the experimental facilities which are necessary for the French naval propulsion program. Since the sixties these facilities have brought a large contribution to the development and to the technical support for the nuclear propulsion; they have been used also to train the French Navy operators. The last experimental reactor, the RNG, is now at the end of its life cycle after thirty years of a profitable operation. A replacement reactor is needed to sustain any evolution of the naval propulsion reactors as well as to guarantee a safe operation and a high level of availability of the existing onboard reactors. The aim of the RES program is namely to build such a test facility. Its construction program started in 2003. By the year 2009 the RES reactor will take over the mission of the RNG. We present hereafter: - A brief history of the French experimental reactors built in support to the naval propulsion, - The needs of the naval propulsion and the related objectives of the RES program, - The corresponding architecture and main characteristics of the RES facility, - The current status of the RES construction. The contents of the paper is as follows: 1. Introduction; 2. History of the French nuclear propulsion experimental reactors; 3. Needs of the naval propulsion and related objectives of the RES reactor; 4. RES architecture and main characteristics; 4.1. The pool module; 4.2. The reactor module; 4.3. The RES reactor, an innovative concept; 5. Realisation status; 6. Conclusion. To summarize, from the year 2009 the RES will be an efficient facility available for irradiation and qualification programs. Its large experimental capabilities will allow relevant fuel and core irradiations. This will give access to a real progress in the knowledge of fuel and core physics as well as in the related simulation tools. This reactor

  17. N Reactor

    Data.gov (United States)

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

  18. Principle of calibration of the simple calorimeter for nuclear heating measurements in MARIA reactor and transposition to the case of JHR reactor

    International Nuclear Information System (INIS)

    The main objective of this project is to develop calculation scheme for the evaluation of gamma heating at any chosen location for both JHR (Jules-Horowitz Reactor) and MARIA reactor. JHR is the future material testing reactor under construction on the Cadarache CEA site (the reactor is going to operate after 2016) and MARIA is an existing Polish reactor having characteristics close to those of JHR. For that purpose both, simulations tools and experimental program, are being developed. Calculations will be carried out with both JHR model and MARIA layout. TRIPOLI4 and APOLLO2 codes will be used for calculations of the gamma heating in reactor cores. Representative calculation scheme for MARIA is needed in view to identify optimal locations for measurements according to JHR needs. The correctness of calculations will be qualified by a comparison with the experimental measurements performed in MARIA reactor. Instrumentation which will be used for these measurements have to be properly calibrated. The main aim of the present paper is to propose a calibration procedure for gamma heating measurements. (authors)

  19. Optimization of coupling desalination systems with a nuclear reactor

    International Nuclear Information System (INIS)

    In 2002, CNSTN (Tunisian National Centre for Nuclear Sciences and Technologies) and CEA-France signed an agreement to jointly pursue, under the aegis of IAEA's Interregional Cooperation Programme INT/4/134, nuclear desalination feasibility studies for the Skhira site in the south of Tunisia. To carry out the technical programme of the agreement, known as the TUNDESAL Project, a mixed team of experts from CEA and from CNSTN as well as from the Tunisian electric utility (STEG) and the water utility (SONEDE) was formed. It was further agreed that the Tunisian experts would partially complete their work at CEA laboratories with IAEA financial assistance (at the Advanced Water Reactor Studies Laboratory, LFEA, of the Innovative Reactor Studies Service, SERI, in the Nuclear Energy Division at the Cadarache Atomic Centre). This feasibility study considers the following tasks: - Pre-dimensioning of the nuclear reactor and desalination processes, compatible with Tunisian electricity needs and required water production capacity at The Skhira site; - Coupling of the selected nuclear reactor to desalination processes and system optimisation; - Economic evaluation of the integrated systems elaborated above; - Safety verification studies of coupled systems. The first three tasks are completed. The fourth task related to safety analysis will be finalized by end 2004

  20. Fuel elements in the core of the reactor Pegase. Description, successive improvements, actual possibilities

    International Nuclear Information System (INIS)

    The core of the research reactor Pegase, in operation at the Cadarache Nuclear Research Centre since 1983, contains fuel elements made from rolled plates of an aluminium-enriched uranium alloy whose characteristics have been changed several times. This report describes the modifications which have been made to these fuel elements with a view both to improving the technical qualities of the reactor and to decreasing its operational costs. Special attention is paid to the neutron aspects of the topic and in particular to the problem of the long-term modification of the reactivity. The 1966 results (30 per cent burn-up associated with only slight movement of the control rods) are particularly satisfying and can probably still be improved in the future. (authors)

  1. The Jules Horowitz Reactor: A New European Material Test Reactor (MTR) Open to International Collaboration: Update Description and Focus on Modern Safety Approach

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new Material Testing Reactor currently under construction at CEA Cadarache research centre in the south of France. It will represent a major Research Infrastructure for scientific studies dealing with material and fuel behaviour under irradiation (and is consequently identified for this purpose within various European road maps and forums; ESFRI, SNE-TP, etc.). The reactor will also be devoted to medical isotopes production. The reactor will perform R and D programs for the optimization of the present generation of NPP, support the development of the next generation of NPP (mainly LWR) and also offer irradiation capacities for future reactors. JHR is designed, built and will be operated as an international user-facility open to international collaboration. In order to comply with the evolution of safety requirements and to guarantee long term operations, the construction safety standards of JHR have been significantly improved compared to MTRs built in the 60s. The paper gives an up-to-date status of the construction and of the developments performed to build the future experimental capacity and is particularly focusing on the modern Safety approach used and its consequences on the design of the reactor. (author)

  2. Reactor Physics

    International Nuclear Information System (INIS)

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  3. Reactor Physics

    International Nuclear Information System (INIS)

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  5. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  6. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation. To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-1, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = deplacement per atom). The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (author)

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

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.1014 ncm-2 s-1 and a fast flux of 6,4.1014 ncm-2s-1, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  9. ITER, the 'Broader Approach', a DEMO fusion reactor

    International Nuclear Information System (INIS)

    Fusion is a very promising future energy option, which is characterized by almost unlimited fuel reserves, favourable safety features and environmental sustainability. The aim of the worldwide fusion research is a fusion power station which imitates the process taking place in the sun and thus gains energy from the fusion of light atomic nuclei. The experimental reactor ITER which will be built in Cadarache, France, marks a breakthrough in the worldwide fusion research: For the first time an energy multiplication factor of at least 10 will be achieved, the factor by which the fusion power exceeds the external plasma heating. Partners in this project are the European Union, Japan, the Russian Federation, USA, China, South Korea and India as well as Brazil as associated partner. The facility is supposed to demonstrate a long burning, reactor-typical plasma and to test techniques such as plasma heating, plasma confinement by superconducting magnets, fuel cycle as well as energy transition, tritium breeding and remote handling technologies. The next step beyond ITER will be the demonstration power station DEMO which requires further developments in order to create the basis for its design and construction. The roadmap to fusion energy is described. It consists of several elements which are needed to develop the knowledge required for a commercial fusion reactor. The DEMO time schedule depends on the efforts in terms of personnel and budget resources the society is willing to invest in fusion taking into account the long term energy supply and its environmental impact. (orig.)

  10. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-09-01

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

  11. A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Pegonen, R., E-mail: pegonen@kth.se [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden); Bourdon, S.; Gonnier, C. [CEA, DEN, DER, SRJH, CEA Cadarache, 13108 Saint-Paul-lez-Durance Cedex (France); Anglart, H. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-10691 Stockholm (Sweden)

    2014-12-15

    Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested.

  12. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  13. Research reactors

    International Nuclear Information System (INIS)

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

  14. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

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

  15. Research reactors

    International Nuclear Information System (INIS)

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world's research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted

  16. Reactor container

    International Nuclear Information System (INIS)

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  17. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Science.gov (United States)

    Kotiluoto, Petri; Wasastjerna, Frej; Kekki, Tommi; Sipilä, Heikki; Banzuzi, Kukka; Kinnunen, Petri; Heikinheimo, Liisa

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  18. Design earthquakes for ITER in Europe at Cadarache. Comparison of the semi-empirical methods used according to the French regulation and the seismic probabilistic assessment of the site

    International Nuclear Information System (INIS)

    Input requirements and assumptions for ITER consider that an infrequent, severe earthquake (called SL-2) which, although unlikely to occur during the lifetime of the facility, is assessed to demonstrate adequate protection of the public. This earthquake is assumed to have a return period of 10,000 years. An investment protection level or inspection level (where all Structures, Systems and Components are safe) with a peak ground acceleration (pga) at 0.5 m/s2 is also considered. As a basis, orders of magnitude of consequences, if no countermeasures were taken, are given. The European site proposed for ITER is situated in the South of France, 40 km North-East of Aix-en-Provence, in a low to moderate seismic area according to the Global Seismic Hazard Map (GSHAP Group 1999). The-tokamak building would be implemented on good bedrock made of limestone with a shear wave velocity of over 1,300 m/s. Four aspects are discussed: regulation, implementation of this regulation for the proposed site (site geology, tectonic, seismo-tectonic), a probabilistic seismic hazard assessment of the site has been performed using a methodology which considers uncertainties (this work has been developed by the team which had contributed to large parts of the GSHAP map study), finally the fulfillment of the requirements and assumptions are discussed, according to IAEA guides for instance. As a conclusion of the studies the main characteristics of the Cadarache European site are discussed: as a reference the standardized 475 year mean return period hazard (i.e. 10 % of chance to occur in 50 years); 10,000 year return period in accordance with the ITER assumption. Preliminary studies have shown that the European site proposal will ensure a low level of project risk with respect to the seismic hazard. (authors)

  19. Reactor building

    International Nuclear Information System (INIS)

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

  20. Heterogeneous reactors

    International Nuclear Information System (INIS)

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author)

  1. OSMOSE program : statistical review of oscillation measurements in the MINERVE reactor R1-UO2 configuration.

    Energy Technology Data Exchange (ETDEWEB)

    Stoven, G.; Klann, R.; Zhong, Z.; Nuclear Engineering Division

    2007-08-28

    The OSMOSE program is a collaboration on reactor physics experiments between the United States Department of Energy and the France Commissariat Energie Atomique. At the working level, it is a collaborative effort between the Argonne National Laboratory and the CEA Cadarache Research Center. The objective of this program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs, and to provide the experimental data for improving the basic nuclear data files. The main outcome of the OSMOSE measurement program will be an experimental database of reactivity-worth measurements in different neutron spectra for the heavy nuclides. This database can then be used as a benchmark to verify and validate reactor analysis codes. The OSMOSE program (Oscillation in Minerve of isotopes in Eupraxic Spectra) aims at improving neutronic predictions of advanced nuclear fuels through oscillation measurements in the MINERVE facility on samples containing the following separated actinides: {sup 232}Th, {sup 233}U, {sup 234}U, {sup 235}U, {sup 236}U, {sup 238}U, {sup 237}Np, {sup 238}Pu, {sup 239}Pu, {sup 240}Pu, {sup 241}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, and {sup 245}Cm. The first part of this report provides an overview of the experimental protocol and the typical processing of a series of experimental results which is currently performed at CEA-Cadarache. In the second part of the report, improvements to this technique are presented, as well as the program that was created to process oscillation measurement results from the MINERVE facility in the future.

  2. Plasma reactor

    OpenAIRE

    Molina Mansilla, Ricardo; Erra Serrabasa, Pilar; Bertrán Serra, Enric

    2008-01-01

    [EN] A plasma reactor that can operate in a wide pressure range, from vacuum and low pressures to atmospheric pressure and higher pressures. The plasma reactor is also able to regulate other important settings and can be used for processing a wide range of different samples, such as relatively large samples or samples with rough surfaces.

  3. Reactor physics

    International Nuclear Information System (INIS)

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  4. Compact Reactor

    International Nuclear Information System (INIS)

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

  5. Jules Horowitz Reactor: Organisation for the Preparation of the Commissioning Phase and Normal Operation

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, J.; Fabre, J. L.; Marcille, O. [French Alternative Energies end Atomic Energy Commission, Provence (France)

    2013-07-01

    The Jules Horowitz Reactor (JHR) is a new modern Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactors design. It will represent also an important research infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which started in 2007 is going-on with target of commissioning by the end of 2017. The design of the reactor provides modern experimental capacity in support to R and D programs for the nuclear energy for the next 60 years. In parallel to the facility construction, the preparation of the future staff and of the organisation to operate the reactor safely, reliably and efficiently is an important issue. In this framework, many actions are in progress to elaborate: Ο the staffing and the organisational structure for the commissioning test phases and also for normal operation, Ο the documentation in support to the reactor operation (safety analysis report, general operating rules, procedures, instructions, ···), Ο the maintenance, in service and periodic test programs, Ο staff training programs by using dedicated facilities (simulator, ···) Ο commissioning test programs for ensuring that the layout of systems and subcomponents is completed in accordance with the design requirements, the specification performances and the safety criteria. These commissioning tests will also be helpful for transferring the knowledge on the installed systems to the operating group. This paper gives the

  6. Jules Horowitz Reactor: Organisation for the Preparation of the Commissioning Phase and Normal Operation

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a new modern Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactors design. It will represent also an important research infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which started in 2007 is going-on with target of commissioning by the end of 2017. The design of the reactor provides modern experimental capacity in support to R and D programs for the nuclear energy for the next 60 years. In parallel to the facility construction, the preparation of the future staff and of the organisation to operate the reactor safely, reliably and efficiently is an important issue. In this framework, many actions are in progress to elaborate: Ο the staffing and the organisational structure for the commissioning test phases and also for normal operation, Ο the documentation in support to the reactor operation (safety analysis report, general operating rules, procedures, instructions, ···), Ο the maintenance, in service and periodic test programs, Ο staff training programs by using dedicated facilities (simulator, ···) Ο commissioning test programs for ensuring that the layout of systems and subcomponents is completed in accordance with the design requirements, the specification performances and the safety criteria. These commissioning tests will also be helpful for transferring the knowledge on the installed systems to the operating group. This paper gives the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Advanced calculational methods for power reactors and LWR core design parameters

    International Nuclear Information System (INIS)

    The purpose of the Specialists Meeting on Advanced Calculational Methods for Power Reactors, held in Cadarache, France, 10-14 September 1990, was to provide a forum for reviewing and discussing selected core physics of water cooled reactors (including high convertors). New methods of advanced calculation for advanced fuels and complex geometries of next generation reactors with a high level of accuracy were discussed and the importance of supercomputing and on-line monitoring was also acknowledged. The meeting was attended by about 60 participants from 20 countries who presented 30 papers. The Technical Committee Meeting on LWR Core Design Parameters, held in Rez, former Czechoslovakia, 7-11 October 1991, provided an opportunity for participants to exchange their experience on reactor physics aspects of benchmark calculations of various lattices, methods for core parameter calculations, core monitoring and in-core fuel management. At the Workshop there were further discussions related to the benchmark problems, homogenization techniques and cross-section representations. Thirty-five papers were presented by about 43 participants from 19 countries. A separate abstract was prepared for each of the mentioned papers. Refs, figs and tabs

  10. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

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

  11. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  13. Thermohydraulic tests of 3x3-rod bundle maquette

    International Nuclear Information System (INIS)

    The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)

  14. Présentation de la Maquette Retraites MARS-2003

    OpenAIRE

    Benoit Bellone; Alexandre Vincent

    2004-01-01

    This paper describes the model MARS, which is the shared property of the DP and the DSS. Currently, MARS aims at planning the general evolution of the first pillar pension scheme for private sector wage-earners, based on macroeconomic and demographic data. It allows to evaluate the impact of changing the parameters used when calculating one’s annuities, or changing general hypotheses. The model provides results which deal with the dynamics of the average annuity and the financial sustainabili...

  15. Multifunctional reactors

    OpenAIRE

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much emphasis in research in the last decade. A survey is given of modern developments and the first successful applications on a large scale. It is explained why their application in many instances is ...

  16. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

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

  17. Nuclear reactor

    International Nuclear Information System (INIS)

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

  18. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations

    International Nuclear Information System (INIS)

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors)

  19. Experimental estimation of the delayed neutron fraction βeff of the MAESTRO core in the MINERVE zero power reactor

    International Nuclear Information System (INIS)

    A method for determining the effective delayed neutron fraction βeff using in-pile reactivity oscillations and Fourier analysis is presented. This method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. This approach enables the estimation of βeff using multi-parameter nonlinear weighted least-squares fit. The method extends previous works by accounting for higher harmonics excitation in the frequency domain by the trapezoidal reactivity signal, both in the reactivity perturbation and in the reactor power response. We show that by using this new approach it is possible to obtain the reactor transfer function in a wide range of frequencies, using only a single oscillation frequency. This method is applied to a set of measurements of the MAESTRO core configuration in the MINERVE zero power reactor (ZPR) located at the Cadarache Research Center. The derived value of βeff, using this method, is 711 ± 17 pcm. (author)

  20. Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

    International Nuclear Information System (INIS)

    Highlights: ► A 1D design approach of a pulsed DEMO reactor is presented. ► The main CS and TF conductor design criteria are presented. ► A typical major radius for a 2 GW DEMO is 9 m. ► A typical plasma magnetic field is 4.9 T. ► The pulse duration is 1.85 h for an aspect ratio of 3. -- Abstract: A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified

  1. Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duchateau, J.-L., E-mail: jean-luc.duchateau@cea.fr [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France); Hertout, P.; Saoutic, B.; Magaud, P.; Artaud, J.-F.; Giruzzi, G.; Bucalossi, J.; Johner, J.; Sardain, P.; Imbeaux, F.; Ané, J.-M.; Li-Puma, A. [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France)

    2013-10-15

    Highlights: ► A 1D design approach of a pulsed DEMO reactor is presented. ► The main CS and TF conductor design criteria are presented. ► A typical major radius for a 2 GW DEMO is 9 m. ► A typical plasma magnetic field is 4.9 T. ► The pulse duration is 1.85 h for an aspect ratio of 3. -- Abstract: A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.

  2. Research reactors - an overview

    Energy Technology Data Exchange (ETDEWEB)

    West, C.D.

    1997-03-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

  3. Safety requirements applied to research reactors in France

    International Nuclear Information System (INIS)

    , the safety requirements in this case are the resistance and leaktightness of the pool and the resistance of the containment building to the water column and overpressure due to this accident. A number of tests were performed to validate the safety demonstrations relating to these issues for the above-mentioned reactors. These tests were particularly aimed at: checking the mechanical strength of the pool structures in the event of a BORAX accident; determining the operating limits of the fuels used; examining the impact on the core of a pressurized irradiation loop rupture; validating the design of clad failure detection systems. The safety reviews systematically performed at the request of the safety authority, for reactors that have been in operation for more than 10 years have not called into question the above-mentioned safety requirements. The main requests of the safety authority following the assessments made by the IRSN and presented to the standing group of experts for the nuclear reactors have concerned the need for upgrading the facilities to take into account in particular the evolutions of the requirements relating to internal and external hazards. In particular, this has led to facilities being reinforced to increase their earthquake resistance, and to improvements concerning fire protection and the management of post-accident situations. The aging of facilities has been addressed by gradually refurbishing their equipment and replacing their instrumentation, their control and protection systems and those structures affected by high radiation doses. The operating experience feedback acquired from existing research reactors, the evolution of certain safety requirements, and the progress achieved in the methodology adopted for the study of research reactor safety have been taken into account in the design of the new RJH reactor (P = 100 MW) which is planned to be constructed in the Cadarache site. The final paper will provide the detailed presentation of the

  4. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

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

    International Nuclear Information System (INIS)

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

  6. Migration behaviour of fission products in and from spherical HTR fuel elements (SAPHIR loop experiments in the PEGASE reactor)

    International Nuclear Information System (INIS)

    The diffusion behaviour of some metallic fission products was evaluated for the irradiation experiments SAPHIR 4-10 and 11 performed by KFA Juelich in the PEGASE reactor, Cadarache (France). Diffusion coefficients for cesium in PyC coatings and for cesium and silver in graphitic fuel element matrix were obtained by analysing the fission product concentration profiles. The reason for the observed different diffusion behaviour of Cs 134 and Cs 137 is discussed as well as that for elevated concentrations of cesium, silver, and ruthenium near the fuel element surface. A simple model for assessing the trapping of metals in a graphite matrix is introduced. Release from the fuel element surfaces into the surrounding gas was found to be governed by direct recoil at low temperatures. (orig.)

  7. Reactor utilization

    International Nuclear Information System (INIS)

    In 1962, the RA reactor was operated almost three times more than in 1961, producing total of 25 555 MWh. Diagram containing comparative data about reactor operation for 1960, 1961, and 1962, percent of fuel used and U-235 burnup shows increase in reactor operation. Number of samples irradiated was 659, number of experiments done was 16. mean powered level was 5.93 MW. Fuel was added into the core twice during the reporting year. In fact the core was increased from 56 to 68 fuel channels and later to 84 fuel channels. Fuel was added to the core when the reactivity worth decreased to the minimum operation level due to burnup. In addition to this 5 central fuel channels were exchanged with fresh fuel in february for the purpose of irradiation in the VISA-2 channel

  8. Reactor Neutrinos

    CERN Document Server

    Lasserre, T; Lasserre, Thierry; Sobel, Henry W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrino oscillation physics in the last years. It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the last undetermined neutrino mixing angle theta13. We conclude by opening on possible use of neutrinos for Society: NonProliferation of Nuclear materials and Geophysics.

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  10. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  11. MINERVE reactor characterization in support of the OSMOSE program: Spectral indices

    International Nuclear Information System (INIS)

    An ambitious program between the Commissariat a l'Energie Atomique (CEA) and the U.S. Dept. of Energy (DOE) has been launched with the aim of measuring the integral absorption rate parameters in the MINERVE experimental facility located at the CEA Cadarache Research Center. The OSMOSE Program (Oscillation in Minerve of isotopes in 'Eupraxic' Spectra) includes a complete analytical program associated with the experimental measurement program and aims at understanding and resolving potential discrepancies between calculated and measured values. The objective of the OSMOSE program is to measure very accurate integral reaction rates in representative spectra for the actinides important to future nuclear system designs and to provide the experimental data for improving the basic nuclear data files. To gain the most information and insight from integral measurements, the essential reactor parameters should be as well characterized as possible. A description of the neutron spectra, or spectral indices, can be obtained via the use of fission chambers of different actinide isotopes. As part of the OSMOSE program, extensive measurements have been performed to characterize the MINERVE reactor. The description and results of the spectral indices measurements of the R1-UO2 and R1-MOX reactor configurations are contained herein. Good agreement was obtained between the experimental measurements and calculations performed using the MCNP Monte Carlo code. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  14. Test devices in Jules Horowitz Reactor dedicated to the material studies in support to the current and future nuclear power plants

    International Nuclear Information System (INIS)

    The Jules Horowitz Reactor (JHR) is a tank pool Material Testing Reactor with a maximum thermal power designed at 100 MW. JHR is being built in the CEA Cadarache site and will take over the Osiris reactor whose decommissioning is planned. JHR's design allows a large experimental capability (around 20 experiments at the same time) inside the reactor core, close to the fuel with high fast neutron flux or outside the reactor core, in the reflector with higher thermal neutron flux. A special attention has been put on the improvement of the thermal stability and gradients of the interest zones in samples despite strong gamma heating and on an improvement of the instrumentation devoted to the experiments. This paper presents the JHR and its main experimental devices that include the MICA (Material Irradiation Capsule) device, the CALIPSO (in-Core Advanced Loop for Irradiation in Potassium and Sodium) loop, the OCCITANE (Out-of-Core Capsule for Irradiation Testing of Ageing by Neutrons) rig, and the CLOE (Corrosion Loop Experiment) loop. JHR will play an important role for Gen IV reactors: CEA studies the feasibility of transmutation capsules, of metal liquid irradiation loops for JHR

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  16. Teaching sodium fast reactor technology and operation for the present and future generations of SFR users

    International Nuclear Information System (INIS)

    This paper provides a description of the education and training activities related to sodium fast reactors, carried out respectively in the French Sodium and Liquid Metal School (ESML) created in 1975 and located in France (at the CEA Cadarache Research Centre), in the Fast Reactor Operation and Safety School (FROSS) created in 2005 at the Phenix plant, and in the Institut National des Sciences et Techniques Nucleaires (INSTN). It presents their recent developments and the current collaborations throughout the world with some other nuclear organizations and industrial companies. Owing to these three entities, CEA provides education and training sessions for students, researchers, and operators involved in the operation or development of sodium fast reactors and related experimental facilities. The sum of courses provided by CEA through its sodium school, FROSS, and INSTN organizations is a unique valuable amount of knowledge on sodium fast reactor design, technology, safety and operation experience, decommissioning aspects and practical exercises. It is provided for the national demand and, since the last ten years, it is extensively opened to foreign countries. Over more than 35 years, the ESML, FROSS, and INSTN have demonstrated their flexibility in adapting their courses to the changing demand in the sodium fast reactor field, operation of PHENIX and SUPERPHENIX plants, and decommissioning and dismantling operations. The results of this ambitious and constant strategy are first sharing of knowledge obtained from experimental studies carried out in research laboratories and operational feedback from reactors, secondly standardized information on safety, and finally the creation of a 'sodium community' that debates, shares the knowledge, and suggests new tracks for a better definition of design and operating rules. (author)

  17. From fundamental mode to the PWR type reactors blow off: physical analysis and contribution to the qualification of calculation tools

    International Nuclear Information System (INIS)

    The work reported in this thesis centres on the resolution of reactor physics problems posed by the use in pressurised water reactors of fuel assemblies containing mixed uranium-plutonium oxide fuel (MOX). The work is essentially dependent on the results of the EPICURE experimental programme carried out between 1988 and 1994 in the reactor EOLE at the Cadarache Research Centre of the CEA. Our contribution to the validation of the computer program APOLLO2 and of its nuclear data library CEA93 shows that this code system satisfactorily calculates the neutronic characteristics of PWR cores. The validation of the experiments has provided useful information concerning the modifications required to be made to the library CEA93, which is based on the basic library of evaluated nuclear data, JEF2. This approach should now be extended to a wider basis of reactor experimental data. The studies of methods for calculating coolant voiding coefficients has made it possible to select suitable methods based on the available deterministic methods of transport theory in 2 ad 3 dimensions. These schemes have given results in satisfactory agreement with the measurements made in EPICURE programme for both local and total coolant voiding. It would now be worth while to validate the chosen methods by comparisons with calculations made using continuous energy Monte Carlo methods. (author)

  18. Reactor container

    International Nuclear Information System (INIS)

    A reactor container has a suppression chamber partitioned by concrete side walls, a reactor pedestal and a diaphragm floor. A plurality of partitioning walls are disposed in circumferential direction each at an interval inside the suppression chamber, so that independent chambers in a state being divided into plurality are formed inside the suppression chamber. The partition walls are formed from the bottom portion of the suppression chamber up to the diaphragm floor to isolate pool water in a divided state. Operation platforms are formed above the suppression chamber and connected to an access port. Upon conducting maintenance, inspection or repairing, a pump is disposed in the independent chamber to transfer pool water therein to one or a plurality of other independent chambers to make it vacant. (I.N.)

  19. Reactor building

    International Nuclear Information System (INIS)

    The present invention concerns a structure of ABWR-type reactor buildings, which can increase the capacity of a spent fuel storage area at a low cost and improved earthquake proofness. In the reactor building, the floor of a spent fuel pool is made flat, and a depth of the pool water satisfying requirement for shielding is ensured. In addition, a depth of pool water is also maintained for a equipment provisionally storing pool for storing spent fuels, and a capacity for a spent fuel storage area is increased by utilizing surplus space of the equipment provisionally storing pool. Since the flattened floor of the spent fuel pool is flushed with the floor of the equipment provisionally storing pool, transfer of horizontal loads applied to the building upon occurrence of earthquakes is made smooth, to improve earthquake proofness of the building. (T.M.)

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  1. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  4. IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility

    International Nuclear Information System (INIS)

    1 - Description: The RRR-SEG-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactor RRG-SEG (at RC Rossendorf / Germany) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in Note Technique SPRC/LEPh/93-230 (SEG) The SEG experiments are considered 'clean' integral experiments, simple and clear in geometry and well suited for calculation. In all SEG configurations only a few materials were used, most of these were standards. Due to the designed adjoint function (energy-independent or monotonously rising), the capture or scattering effect was dominant, convenient to check separately capture or scattering data. At first, analyses of the experiments have been performed in Rossendorf. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and

  5. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  6. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

  8. Nuclear research reactors

    International Nuclear Information System (INIS)

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.)

  9. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  10. Sensitivity and uncertainty analysis on reactor kinetic parameters using perturbation theory

    International Nuclear Information System (INIS)

    The analysis of nuclear reactors behavior in transient depends on the neutronic parameters including the effective delayed neutron fraction (βeff) and its family dependent components (βeff), the neutron generation lifetime (lp), and the decay constants (λi) for each delayed neutron family. With the aim of an accurate assessment of neutron parameters used to simulate transient, uncertainties on these parameters must be estimated by considering the contributions of different nuclear data, including those from the delayed data (fission yield νd and fission spectrum χd). This paper presents a methodology based on the generalized perturbation theory which allows the assessment of kinetic parameters sensitivities to nuclear data. The methodology is applied to the heterogeneous core design 'CFV' considered for sodium-cooled reactor ASTRID. The contributions of indirect term and direct terms (σfission ν and νd , χ and χd) are decomposed by isotopes and reactions and uncertainties on the kinetic parameters are calculated using the most recent data covariance suggested in COMAC (COvariance Matrices from Cadarache, version 0.1). (author)

  11. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Science.gov (United States)

    Fourmentel, D.; Filliatre, P.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Carcreff, H.

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g-1 and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

  13. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  14. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  15. Reactor container

    International Nuclear Information System (INIS)

    Purpose: To prevent shocks exerted on a vent head due to pool-swell caused within a pressure suppression chamber (disposed in a torus configuration around the dry well) upon loss of coolant accident in BWR type reactors. Constitution: The following relationship is established between the volume V (m3) of a dry well and the ruptured opening area A (m2) at the boundary expected upon loss of coolant accident: V >= 30340 (m) x A Then, the volume of the dry well is made larger than the ruptured open area, that is, the steam flow rate of leaking coolants upon loss of coolant accident to decrease the pressure rise in the dry well at the initial state where loss of coolant accident is resulted. Accordingly, the pressure of non-compressive gases jetted out from the lower end of the downcomer to the pool water is decreased to suppress the pool-swell. (Ikeda, J.)

  16. Research reactor status for future nuclear research in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Raymond, Patrick; Bignan, Gilles; Guidez, Joel [Commissariat a l' Energie Atomique - CEA (France)

    2010-07-01

    construction on the Cadarache site. Other projects as PALLAS, MYRRHA, ASTRID are described. We have to pinpoint the fact that these research reactors are more and more international reactors used as international user's platform. Now 13 counties are operating the ILL in Grenoble, and for the future JHR under construction, 10 countries are now part of the consortium. (authors)

  17. Irradiation in the CAP reactor of Fragema fuel assemblies under power transients representative of super imposed load follow and remote control regimes

    International Nuclear Information System (INIS)

    The CAP (Chaufferie Avancee Prototype) is a pressurized research reactor of about 100 MWth located at the Nuclear Center of Cadarache. The first core with fuel rods was loaded in 1976; in 1978, four fuel assemblies of current design for the French PWRs were loaded in the second core; the objective of this last experiment was to test the reference fuel under superimposed daily load follow (power reduction to 75%, 50% or 30% of the rated power during 6 hours) and remote control (power level change of approximately ±5% regimes. The rated power of the CAP reactor was selected to give the same rod linear ratings as in a 900 MW.PWR. The local changes in power during a reactor transient had been previously analyzed and reproduced in the CAP reactor by the motion of the control rods and power changes. The campaign was successfully carried out without any fuel problem and, after unloading the core, the sipping-tests and subsequent PIE did not show any cladding failure. The four fuel assemblies were loaded in the third core and two of them will reach a mean burn-up of about 33,000 MWd/tU at the end of this campaign, by mid 1983. These irradiations provide representative information for the evaluation of fuel behaviour under severe power cycling conditions and are completed by non-destructive and destructive examinations and tests performed on characterized fuel rods removed from the assemblies at the end of each cycle. (author)

  18. Third research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors'. Working material

    International Nuclear Information System (INIS)

    The general objective of the Co-ordinated Research Project (CRP) on Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFBR Reactivity Effect, is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first Research Co-ordination Meeting (RCM) took place in Vienna on 24-26 November 1999 and the third RCM was scheduled for 12-16 November 2001 in Cadarache France. The objectives of the third RCM were: to review and finalize the draft synthesis report of Phase I (R-Z Homogeneous reactor core calculations) and Phase 2 (Hex-Z Homogeneous reactor core calculations); to compare the results of Phase 3 (Hex-Z Heterogeneous reactor core calculations and Burnup); to agree on the work scope of Phase 4 (BFS-62 Analysis); to discuss proposals for Phase 5 (Minor Actinides). In this context, review and related discussions were made on the following items: summary review of Actions and results since last meeting; presentation of individual results for Phase 3 by Member States; discussion of the comparison of the results for Phases 1 and 2; preliminary analysis of the results for Phase 3; definition of the work to be performed under Phase 4 of CRP; discussion of work scope for the next CRP Phases

  19. Validation of iron nuclear data for the neutron calculation of nuclear reactors

    International Nuclear Information System (INIS)

    The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of 56Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the 56Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an iterative technique

  20. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  1. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  2. RB reactor noise analysis

    International Nuclear Information System (INIS)

    Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed

  3. Needs of Accurate Prompt and Delayed γ-spectrum and Multiplicity for Nuclear Reactor Designs

    Science.gov (United States)

    Rimpault, G.; Bernard, D.; Blanchet, D.; Vaglio-Gaudard, C.; Ravaux, S.; Santamarina, A.

    The local energy photon deposit must be accounted accurately for Gen-IV fast reactors, advanced light-water nuclear reactors (Gen-III+) and the new experimental Jules Horowitz Reactor (JHR). The γ energy accounts for about 10% of the total energy released in the core of a thermal or fast reactor. The γ-energy release is much greater in the core of the reactor than in its structural sub-assemblies (such as reflector, control rod followers, dummy sub-assemblies). However, because of the propagation of γ from the core regions to the neighboring fuel-free assemblies, the contribution of γ energy to the total heating can be dominant. For reasons related to their performance, power reactors require a 7.5% (1σ) uncertainty for the energy deposition in non-fuelled zones. For the JHR material-testing reactor, a 5% (1 s) uncertainty is required in experimental positions. In order to verify the adequacy of the calculation of γ-heating, TLD and γ-fission chambers were used to derive the experimental heating values. Experimental programs were and are still conducted in different Cadarache facilities such as MASURCA (for SFR), MINERVE and EOLE (for JHR and Gen-III+ reactors). The comparison of calculated and measured γ-heating values shows an underestimation in all experimental programs indicating that for the most γ-production data from 239Pu in current nuclear-data libraries is highly suspicious.The first evaluation priority is for prompt γ-multiplicity for U and Pu fission but similar values for otheractinides such as Pu and U are also required. The nuclear data library JEFF3.1.1 contains most of the photon production data. However, there are some nuclei for which there are missing or erroneous data which need to be completed or modified. A review of the data available shows a lack of measurements for conducting serious evaluation efforts. New measurements are needed to guide new evaluation efforts which benefit from consolidated modeling techniques.

  4. Research Nuclear Reactors

    International Nuclear Information System (INIS)

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

  5. Temperature effects studies in light water reactor lattices

    International Nuclear Information System (INIS)

    The CREOLE experiments performed in the EOLE critical facility located in the Nuclear Center of CADARACHE - CEA (UO2 and UO2-PuO2 lattice reactivity temperature coefficient continuous measurements between 200C and 3000C; integral measurements by boron equivalent effect in the moderator; water density effects measurements with the use of over cladding aluminium tubes to remove moderator) allow to get an interesting and complete information on the temperature effects in the light water reactor lattices. A very elaborated calcurated scheme using the transport theory and the APOLLO cross sections library, has been developed. The analysed results of the whole lot of experiments show that the discrepancy between theory and experiment strongly depends on the temperature range and on the type of lattices considered. The error is mainly linked with the thermal spectrum effects. A study on the temperature coefficient sensitivity to the different cell neutron parameters has shown that only the shapes of the 235U and 238U thermal cross sections have enough weight and uncertainty margins to explain the observed experimental/calculation bias. Instead of arbitrarily fitting the identified wrong data on the calculation of the reactivity temperature coefficient we have defined a procedure of modification of the cross sections based on the consideration of the basic nuclear data: resonance parameters and associated statistic laws. The implementation of this procedure has led to propose new thermal cross sections sets for 235U and 238U consistent with the uncertainty margins associated with the previously accepted values and with some experimental data

  6. Reactor Physics Training

    International Nuclear Information System (INIS)

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  7. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

  8. Safeguarding research reactors

    International Nuclear Information System (INIS)

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  9. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  10. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch

    2008-01-15

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  11. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  12. Neutron Beam Analysis on Materials for Nuclear Applications, Being Irradiated in Fission Reactors and Having Radioactivity

    International Nuclear Information System (INIS)

    Extensive supports are given from the public sectors to the neutron beam analysis on advanced materials developed mainly in the framework of fundamental solid state physics, through the Japan Atomic Energy Agency and the Institute for Solid State physics in University of Tokyo. However, the related activities are mainly on non-radioactive materials with some limited exceptions, though the facilities for the neutron beam analysis are installed in the radiation controlled areas. Research activities in the field of nuclear related materials have concentrated their efforts for nano structural analysis into the other techniques of the post irradiation examinations, such as the high resolution transmission microscopy, the three dimensional atom probe tomography, and the positron annihilation techniques, than the neutron beam analysis. In the meantime, more detailed analysis on the radiation induced nanostructures are becoming more and more essential for the further understanding of the radiation effects in the materials which will be used in the advanced nuclear systems, such as the nuclear fusion reactors and the generation-IV nuclear fission reactors. Utilizing of the cutting edge techniques for the nanostructural analysis on materials irradiated by neutrons, all of which cannot be installed in the limited area of available hot laboratories, is urgently demanded, of course, satisfying the related legal restrictions and the safety demands. The present study was focused on as the realization of the neutron beam analysis on the nanostructural evolutions of the superconductive materials, which will be used in the ITER, the international thermonuclear experimental reactor, being under construction in Cadarache, France, and the glassy metals, which have some unique and advantageous features for the nuclear applications. (author)

  13. Conceptual integrated approach for the magnet system of a tokamak reactor

    International Nuclear Information System (INIS)

    Highlights: • We give a conceptual approach of a fusion reactor magnet system based on analytical formula. • We give design criteria for the CS and TF cable in conduit conductors and for the magnet system structural description. • We apply this conceptual approach to ITER and we crosscheck with actual characteristics. • We apply this conceptual approach to a possible version of DEMO. - Abstract: In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO

  14. Development and qualification of reference calculation schemes for absorbers in pressured water reactor

    International Nuclear Information System (INIS)

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code APOLLO2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI4. They were then checked against experimental data measured during French experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  15. Elaboration and qualification of a reference calculation routes for the absorbers in the PWR reactors

    International Nuclear Information System (INIS)

    The general field in which this work takes place is the field of the accuracy improvement of neutronic calculations, required to operate Pressurized Water Reactors (PWR) with a better precision and a lower cost. More specifically, this thesis deals with the calculation of the absorber clusters used to control these reactors. The first aim of that work was to define and validate a reference calculation route of such an absorber cluster, based on the deterministic code Apollo 2. This calculation scheme was then to be checked against experimental data. This study of the complex situation of absorber clusters required several intermediate studies, of simpler problems, such as the study of fuel rods lattices and the study of single absorber rods (B4C, AIC, Hafnium) isolated in such lattices. Each one of these different studies led to a particular reference calculation route. All these calculation routes were developed against reference continuous energy Monte-Carlo calculations, carried out with the stochastic code TRIPOLI14. They were then checked against experimental data measured during french experimental programs, undertaken within the EOLE experimental reactor, at the Nuclear Research Center of Cadarache: the MISTRAL experiments for the study of isolated absorber rods and the EPICURE experiments for the study of absorber clusters. This work led to important improvements in the calculation of isolated absorbers and absorber clusters. The reactivity worth of these clusters in particular, can now be obtained with a great accuracy: the discrepancy observed between the calculated and the experimental values is less than 2.5 %, and then slightly lower than the experimental uncertainty. (author)

  16. Conceptual integrated approach for the magnet system of a tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duchateau, J.-L., E-mail: Jean-luc.duchateau@cea.fr; Hertout, P.; Saoutic, B.; Artaud, J.-F.; Zani, L.; Reux, C.

    2014-11-15

    Highlights: • We give a conceptual approach of a fusion reactor magnet system based on analytical formula. • We give design criteria for the CS and TF cable in conduit conductors and for the magnet system structural description. • We apply this conceptual approach to ITER and we crosscheck with actual characteristics. • We apply this conceptual approach to a possible version of DEMO. - Abstract: In the framework of the reflexion about DEMO, a conceptual integrated approach for the magnet system of a tokamak reactor is presented. This objective is reached using analytical formulas which are presented in this paper, coupled to a Fortran code ESCORT (Electromagnetic Superconducting System for the Computation of Research Tokamaks), to be integrated into SYCOMORE, a code for reactor modelling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modelling task force. The analytical formulas deal with all aspects of the magnet system, starting from the derivation of the TF system general geometry, from the plasma main characteristics. The design criteria for the cable current density and the structural design of the toroidal field and central solenoid systems are presented, enabling to deliver the radial thicknesses of the magnets and enabling also to estimate the plasma duration of the plateau. As a matter of fact, a pulsed version DEMO is presently actively considered in the European programmes. Considerations regarding the cryogenics and the protection are given, affecting the general design. An application of the conceptual approach is presented, allowing a comparison between ESCORT output data and actual ITER parameters and giving the main characteristics of a possible version for DEMO.

  17. The use and evolution of the CEA research reactors

    International Nuclear Information System (INIS)

    organic liquids under irradiation. The equipments of Melusine are now being modernized. Siloe, another swimming-pool reactor, has been operating at 15 MW since the end of 1963. The performances achieved constitute a considerable progress in the field of swimming-pool reactors, since the fluxes obtained with Siloe are in the same order of magnitude than those which needed till now a tank type structure, whose numerous disadvantages in the fitting of experiments are well known. Siloe will be used mostly in the study of structural materials, graphites, refractory fissile materials, and for solid state physics. The reactor Pegase, in service in the Cadarache Nuclear Centre since 1963, is intended solely for testing full-scale fuel elements of EDF and EL 4 types. The current programme, for the eight loops of the reactor, covers the elements for the reactors EDF 2, EDF 3 and EL 4. New loops are in the course of being studied for the fuel elements of the EDF 4 and EDF 5 reactors. The general line of the CEA programmes has shown up the considerable need for fast neutron irradiations. The reactor Osiris which is in the course of being constructed will serve to complement the CEA's equipment in this field and at the same time fill the gap which will be left in the near future in the Centre de Saclay by the closing down of EL 2. Osiris is a light water reactor whose special structure will allow it to function at 50 MW without the disadvantages usually associated with the presence of a heavy waterproof tank. This reactor, which should be put into service in 1966, is mainly intended for the investigation of structural materials, graphite and refractory fuels; it will also serve to increase the production of high specific activity isotopes, and to develop activation analysis techniques. (authors)

  18. Reactor Physics Programme

    International Nuclear Information System (INIS)

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  19. Ship propulsion reactors technology

    International Nuclear Information System (INIS)

    This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

  20. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  1. Process heat reactors

    International Nuclear Information System (INIS)

    The consumption of heat, for industrial and domestic needs, takes up half of the national energy supply; direct utilization of the heat produced by nuclear reactors could therefore contribute to reduce the deficit in the energetic results. The restraints proper to heat consumption (dispersal and variety of consumers, irregular demand) involve the development of the heat transport system structures and adequate nuclear reactors. With this in view, the Commissariat a l'Energie Atomique and Technicatome are developing the CAS reactor series, pressurized water reactors (PWR), (CAS 3G reactor with a power of 420 MW.th.), and the Thermos reactor (100 MW.th.), directly conceived to produce heat at 1200C and whose technology derives from the experimental pool reactors type. In order to prove the value of the Thermos design, an experimental reactor should soon be constructed in the Saclay nuclear research centre

  2. Reactor System Design

    International Nuclear Information System (INIS)

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  3. Structural behaviour of fuel assemblies for water cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    At the invitation of the Government of France and in response to a proposal of the IAEA Technical Working Group on Water Reactor Fuel Performance and Technology (TWGFPT), the IAEA convened a Technical Meeting on Fuel Assembly Structural Behaviour in Cadarache, France, from 22 to 26 November 2004. The meeting was hosted by the CEA Cadarache Centre, AREVA Framatome-ANP and Electricite de France. The meeting aimed to provide in depth technical exchanges on PWR and WWER operational experience in the field of fuel assembly mechanical behaviour and the potential impact of future high burnup fuel management on fuel reliability. It addressed in-service experience and remedial solutions, loop testing experience, qualification and damage assessment methods (analytic or experimental ones), mechanical behaviour of the fuel assembly including dynamic and fluid structure interaction aspects, modelling and numerical analysis methods, and impact of the in-service evolution of the structural materials. Sixty-seven participants from 17 countries presented 30 papers in the course of four sessions. The topics covered included the impact of hydraulic loadings on fuel assembly (FA)performance, FA bow and control rod (CR) drop kinetics, vibrations and rod-to-grid wear and fretting, and, finally, evaluation and modelling of accident conditions, mainly from seismic causes. FA bow, CR drop kinetics and hydraulics are of great importance under conditions of higher fuel duties including burnup increase, thermal uprates and longer fuel cycles. Vibrations and rod-to-grid wear and fretting have been identified as a key cause of fuel failure at PWRs during the past several years. The meeting demonstrated that full-scale hydraulic tests and modelling provide sufficient information to develop remedies to increase FA skeleton resistance to hydraulic loads, including seismic ones, vibrations and wear. These proceedings are presented as a book with an attached CD-ROM. The first part of the CD

  4. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  5. The Chernobylsk reactor accident

    International Nuclear Information System (INIS)

    The construction, the safety philosophy, the major reactor physical parameters of RBMK-1000 type reactor units and the detailed description of the Chernobylsk-4 reactor accident, its causes and conclusions, the efforts to reduce the consequences on the reactor site and in the surroundings are discussed based on different types of Soviet documents including the report presented to the IAEA by the Soviet Atomic Energy Agency in August 1986. (V.N.)

  6. Zero energy reactor 'RB'

    International Nuclear Information System (INIS)

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  7. High solids fermentation reactor

    Science.gov (United States)

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  8. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  9. Fusion reactor studies

    International Nuclear Information System (INIS)

    A review is given of fusion reactor systems studies, the objectives of these studies are outlined and some recent conceptual reactor designs are described. The need for further studies in greater depth is indicated so that progress towards a commercial fusion reactor may be consolidated. (U.K.)

  10. Reactor power measuring device

    International Nuclear Information System (INIS)

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  11. Monte Carlo interpretation of the AMMON/REF experiment in EOLE for the JHR reactor neutron calculations

    International Nuclear Information System (INIS)

    A new experimental program, named AMMON, was performed between late 2010 and early 2013 in the EOLE zero-power experimental reactor at CEA Cadarache. It is dedicated to the analysis of the neutron and photon physics of the Jules Horowitz Reactor (JHR), the next international materials testing reactor under construction in France. The objective of the program is to provide measurement data for the experimental validation of the calculation tools developed for the JHR design and safety studies. The first core configuration, the so-called reference configuration, was loaded in 2012; it consisted of an experimental zone of seven JHR assemblies with U3Si2-Al, 27% 235U enriched fuel curved plates surrounded by a driver zone with 622 standard pressurized water reactor uranium oxide fuel pins. It has been instrumented and studied throughout the first year of the experimental program. The final analysis of the AMMON/REF neutron measurements is presented in this paper. It is based on calculations performed with the three-dimensional reference Monte Carlo TRIPOLI-4.7 code and the JEFF3.1.1 European library. The comparison between calculation and experiment makes it possible to calibrate the bias due to nuclear data on the calculated neutron parameters. It highlights good agreement between calculation and experiment concerning reactivity, power distribution in the experimental zone, fuel plate conversion ratios, and core kinetics parameters. The reactivity prediction is very satisfactory, despite the presence of a large aluminum quantity in the core: calculation-to-experiment comparison (C-E=)365 ± 334 pcm (1σ). For the other neutron parameters (assembly power distribution, plate conversion ratios, and kinetics parameters), the (C-E)/E discrepancies are within the experimental uncertainty (2σ). (authors)

  12. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  13. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  14. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  15. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  16. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    While the power generation needs of utilities are unique and diverse, they are all faced with the double challenge of meeting growing electricity needs while curbing CO2 emissions. To answer these diverse needs and help tackle this challenge, AREVA has developed several reactor models which are briefly described in this document: The EPRTM Reactor: designed on the basis of the Konvoi (Germany) and N4 (France) reactors, the EPRTM reactor is an evolutionary model designed to achieve best-in-class safety and operational performance levels. The ATMEA1TM reactor: jointly designed by Mitsubishi Heavy Industries and AREVA through ATMEA, their common company. This reactor design benefits from the competencies and expertise of the two mother companies, which have commissioned close to 130 reactor units. The KERENATM reactor: Designed on the basis of the most recent German BWR reactors (Gundremmingen) the KERENATM reactor relies on proven technology while also including innovative, yet thoroughly tested, features. The optimal combination of active and passive safety systems for a boiling water reactor achieves a very low probability of severe accident

  17. The Maple reactor project

    International Nuclear Information System (INIS)

    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)

  18. High temperature reactors

    International Nuclear Information System (INIS)

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements

  19. Spinning fluids reactor

    Science.gov (United States)

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  20. Reactor Safety: Introduction

    International Nuclear Information System (INIS)

    The programme of the Reactor Safety Division focuses on the development of expertise on materials behaviour under irradiation for fission and fusion oriented applications. Furthermore, as nuclear energy needs international public acceptance with respect to safety and efficient management of natural resources and wants to reduce the burden of nuclear waste, the Reactor Safety Division enhanced its efforts to develop the MYRRHA project. MYRRHA, an accelerator driven sub-critical system, might have the potential to cope in Europe with the above mentioned constraints on acceptability and might serve as a technological platform for GEN IV reactor development, in particular the Liquid Metal Fast Reactor.The Reactor Safety Division gathers three research entities that are internationally recognised: the Reactor Materials Research department, the Reactor Physics and MYRRHA department and the Instrumentation department.The objectives of Reactor Materials Research are: to evaluate the integrity and behaviour of structural materials and nuclear fuels used in present and future nuclear power industry; to perform research to unravel and understand the parameters that determine the material and fuel behaviour under or after irradiation; to contribute to the interpretation and modelling of the materials and fuels behaviour in order to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the Reactor Materials Research department concentrate on four distinct disciplines: Reactor Pressure Vessel Steel embrittlement Stress corrosion cracking in reactor coolant environment, including Irradiation Assisted Stress Corrosion Cracking; Nuclear Fuel characterisation and development of new fuel types for commercial and test reactors. Development of materials for Fusion and advanced nuclear fission reactors. The safe operation of present nuclear power plants relies primarily on the integrity of the reactor pressure vessel

  1. Order of the 9. of March 2010 approving the decision no 2010-DC-0172 of the 5. of January 2010 by the Nuclear Safety Authority specifying the limits of releases in the environment of liquid and gaseous effluents of civil base nuclear installations of the Cadarache Centre operated by the Atomic Energy Commissariat (CEA) on the district of Saint-Paul-lez-Durance (Bouches-du-Rhone department); Arrete du 9 mars 2010 portant homologation de la decision no 2010-DC-0172 de l'Autorite de surete nucleaire du 5 janvier 2010 fixant les limites de rejets dans l'environnement des effluents liquides et gazeux des installations nucleaires de bases civiles du centre de Cadarache exploitees par le Commissariat a l'energie atomique (CEA) sur la commune de Saint-Paul-lez-Durance (departement des Bouches-du-Rhone)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    This legal publication contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents of civil base nuclear installations of the Cadarache Centre operated by the Atomic Energy Commissariat (CEA) on the district of Saint-Paul-lez-Durance. These installations are listed and tables present the limits for different radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and alpha, beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (HCl, HF, NOx, CO, and so on), as well as limits for thermal releases

  2. European Fast Reactor IWGFR/FRCC-report. A review of the collaborative programme on the European Fast Reactor (EFR)

    International Nuclear Information System (INIS)

    The design work for the 1500 MWe European Fast Reactor EFR was started in 1988. Two years during phase 1 were devoted to the concept design; the subsequent concept validation phase 2 will last until March 1993. In autumn 1991 the 'concept design '91, CD91, was put forward; its major design features and the R and D support are described briefly together with the organisational structures. The European Fast Reactor Utilities Group 'EFRUG' presently comprises EdF (France), ENEL (Italy), Nuclear Electric (UK) and Bayernwerk, PreulsenElektra and RWE (Germany). For design and construction of EFR the group 'EFR Associates' is responsible, combining the companies Siemens (formerly Interatom, Germany), NNC Ltd. (UK) and Framatome/Div. Novatome (France). The necessary R and D support is given by CEA (France), UKAEA (UK) and KfK/Siemens (Germany). The R and D work is executed in the various national research centres ranging from Dounreay via Bensberg and Karlsruhe to Cadarache. The design work is done at Bensberg, Lyon and Risley. The present programme of design work extends to early 1993 and is aimed at producing a detailed consistent design for the nuclear part of the plant and a non site specific safety report. By that date the basic feasibility of the main design features will have been underwritten by the joint R and D programme and there will be an informal assessment of the general licensibility of the concept by the Ad Hoc Safety Club. In follow-up the utilities will then be in the position to decide whether to proceed with the next steps. The key issues of this phase will include the specification of the plant, the siting, the detailed engineering, licensing with possibly a public enquiry and the question of ownership and financing. In the international arena the collaboration with USSR is proceeding well on the basis of an USSR-Europe Agreement from January 8, 1991; it foresees review and specialists meetings in the field of fast breeder research. On the occasion of

  3. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  4. Thai research reactor

    International Nuclear Information System (INIS)

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  5. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  6. Safety report concerning the reactor Pegase - volume 1 - Description of the installation - volume 2 - Safety of the installations; Rapport de surete du reacteur pegase - tome 1 - Description des installations - tome 2 - Surete des installations

    Energy Technology Data Exchange (ETDEWEB)

    Lacour, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Legoin, P. [S.E.M. Hispano-Suiza, 92 - Colombes (France)

    1964-07-01

    In the first volume: This report is a description of the reactor Pegase, given with a view to examine the safety of the installations. The Cadarache site at which they are situated is briefly described, in particular because of the consequences on the techniques employed for building Pegase. A description is also given of the original aspects of the reactor. The independent loops which are designed for full-scale testing of fuel elements used in natural uranium-gas-graphite reactor systems are described in this report, together with their operational and control equipment. In the second volume: In the present report are examined the accidents which could cause damage to the Pegase reactor installation. Among possible causes of accidents considered are the seismicity of the region, an excessive power excursion of the reactor and a fracture in the sealing of an independent loop. Although all possible precautions have been taken to offset the effects of such accidents, their ultimate consequences are considered here. The importance is stressed of the security action and regulations which, added to the precautions taken for the construction, ensure the safety of the installations. (authors) [French] Dans le volume 1: Ce rapport est une description du reacteur Pegase, afin d'examiner la surete des installations. Le site de CADARACHE ou elles sont situees, a ete sommairement decrit, en particulier, a cause des consequences sur les techniques mises en oeuvre pour la realisation de Pegase. Nous nous sommes egalement attache a decrire les aspects originaux du reacteur. Les boucles autonomes destinees a tester en vraie grandeur des elements combustibles de la filiere uranium naturel graphite-gaz, ainsi que leurs dispositifs de controle et d'exploitation, figurent egalement dans ce rapport. Dans le volume 2: Dans le present rapport, nous examinons des accidents pouvant endommager des installations du reacteur Pegase. Les origines d'accidents examines

  7. Order of the 9. of March 2010 approving the decision no 2010-DC-0172 of the 5. of January 2010 by the Nuclear Safety Authority specifying the limits of releases in the environment of liquid and gaseous effluents of civil base nuclear installations of the Cadarache Centre operated by the Atomic Energy Commissariat (CEA) on the district of Saint-Paul-lez-Durance (Bouches-du-Rhone department)

    International Nuclear Information System (INIS)

    This legal publication contains references to the different legal and official documents (codes, orders, minister's opinion, public surveys, administrative authorizations, local community opinion) at the root of this specification of limits related to releases in the environment of liquid and gaseous effluents of civil base nuclear installations of the Cadarache Centre operated by the Atomic Energy Commissariat (CEA) on the district of Saint-Paul-lez-Durance. These installations are listed and tables present the limits for different radioactive emissions (carbon 14, hydrogen 3, iodine, radioactive rare earths, and alpha, beta and gamma emitters) from these different installations, but also the maximum admitted concentrations for gaseous chemical effluents (HCl, HF, NOx, CO, and so on), as well as limits for thermal releases

  8. On-line fast flux measurements in the BR2 reactor

    International Nuclear Information System (INIS)

    Since 2001, CEA-Cadarache and the Belgian Nuclear Research Centre SCK-CEN are collaborating on the development and in-pile qualification of subminiature fission chambers (diameter of 1.5 mm). Initially, efforts concentrated on fission chambers for the in-pile measurement of thermal fluxes (with 235U as fissile material). Meanwhile successful long-term tests of the prototypes have been performed in various environments: in low temperature (40-100 degress Celsius) BR2 pool water (up to a thermal neutron fluence of 3 1021 n/cm2) and in the CALLISTO PWR loop (300 degrees Celsius, 155 bars). The long-term qualification of derived industrial detectors (Photonis CFUZ53) in CALLISTO is still ongoing. However, for various types of irradiations in research reactors, the knowledge of the evolution of the fast neutron flux is even of more interest than the thermal flux data. Therefore the collaboration program was extended to the development and the in-pile qualification of subminiature or miniature fission chambers (with 3 mm diameter) for fast neutron detection, for which 242Pu was selected as the optimal fissile material. In order to achieve the on-line in-pile measurement of fast neutron flux, the fission chambers will be operated in the Campbelling mode (based on the mean square fluctuation of the detector current). In this mode the gamma induced contribution to the signal can be efficiently suppressed. Moreover, a data processing software will take into account the evolution of the fissile deposit in order to assess on-line the fast flux sensitivity and to correct for the low energy neutron contributions. The final objective is to qualify a Fast Neutron Detector System (FNDS) able to provide on-line data for local fast neutron fluxes in Material Testing Reactors. The on-line measurement of the fast neutron flux would contribute significantly to the characterization of the irradiation conditions during test experiments with materials and innovative fuel elements

  9. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  10. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  11. Evaluation of research reactors

    International Nuclear Information System (INIS)

    The present status of research reactors with highly enriched (93%) uranium fuel at JAERI, JRR-2 and JMTR is described. JRR-2 is a heterogeneous type of reactor, using heavy water as moderator and coolant. It uses both MTR type and cylindrical type of fuel elements. The maximum thermal power and the thermal neutron flux are 10 MW and 2x1014 n/cm2 see respectively. The reactor has been used for various experiments such as solid state physics, material irradiation, reactor fuel irradiation and radioisotope production. The JMTR is a multi-purpose tank type material testing reactor, and light water moderator and coolant, operated at 50 MW. The evaluation of lower enriched fuel and its consequences for both reactors is considered more especially

  12. Multipurpose research reactors

    International Nuclear Information System (INIS)

    The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  13. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  14. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  15. Phenomenological modeling and study of a catalytic membrane reactor for water detritiation

    International Nuclear Information System (INIS)

    Tritium is produced in light and heavy water reactor fuel by ternary fission or neutron activation. This by-product is used as fuel in fusion fuel reactors such as JET in Culham or ITER in Cadarache (France). The growing interest of this research area will make the tritium fluxes increase; it is then worth addressing the question of its future whether it will be used or flushed out from liquid and gaseous effluents or waste. This thesis studies the recovery of tritium as fuel for fusion machines by means of packed bed membrane reactor (PBMR). Such a reactor combines catalytic conversion of tritiated water thanks to isotope exchange with hydrogen according to the reversible reaction Q2O+H2↔H2O+Q2 (Q=H,D or T) and selective permeation of Q2 through Pd-based membrane. In fact, palladium has the ability to bond with hydrogen isotopes, creating a selective permeation barrier. In the PBMR, thanks to the reaction products withdrawal, these permeation fluxes drive the heavy water conversion rate, to higher values than those reached in conventional fixed bed reactors (Le Chatelier's law). In order to study PBMRs, the CEA has built a test bench, using deuterium instead of tritium, allowing the analysis of their conversion and separation performances at the laboratory scale. An in-house method has been developed to determine simultaneously hydrogen and water isotopologues content by mass spectrometer analysis. It was experimentally shown that the activity of Ni-based catalyst used in this study was sufficient to allow the isotope exchange reactions to reach their thermodynamic equilibrium in a very short time. In addition, hydrogen permeation flux was shown to follow a Richardson's law. Sensitivity studies performed on the PBMR's main operating parameters revealed that its global performance (i.e. de-deuteration factor) increases with the temperature, the transmembrane pressure difference, the sweep gas flow rate and the residence time in the catalyst

  16. Fusion reactor research

    International Nuclear Information System (INIS)

    This work covers four separate areas: (1) development of technology for processing liquid lithium from blankets, (2) investigation of hydrogen isotope permeation in candidate structural metals and alloys for near-term fusion reactors, (3) analytical studies encompassing fusion reactor thermal hydraulics, tritium facility design, and fusion reactor safety, and (4) studies involving dosimetry and damage analysis. Recent accomplishments in each of these areas are summarized

  17. The Integral Fast Reactor

    International Nuclear Information System (INIS)

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab

  18. The replacement research reactor

    International Nuclear Information System (INIS)

    As a consequences of the government decision in September 1997. ANSTO established a replacement research reactor project to manage the procurement of the replacement reactor through the necessary approval, tendering and contract management stages This paper provides an update of the status of the project including the completion of the Environmental Impact Statement. Prequalification and Public Works Committee processes. The aims of the project, management organisation, reactor type and expected capabilities are also described

  19. PFBR reactor protection

    International Nuclear Information System (INIS)

    Design philosophy adopted for Prototype Fast breeder Reactor (PFBR) is a classical one and has the following features: triplicated sensors for measuring important safety parameters; two independent reactor protection Logic Systems based on solid state devices; reactivity control achieved by control rods; gas equipped modules at the core blanket interface providing negative reactivity. Design verification of these features showed that safety of the reactor can be achieved by a traditional approach since the inherent features of LMFBR make this easy

  20. Reactor BR2

    International Nuclear Information System (INIS)

    The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

  1. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  2. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  3. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  4. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  5. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  6. New reactor concepts

    International Nuclear Information System (INIS)

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  7. Reactor construction steels

    International Nuclear Information System (INIS)

    The basic functions of light water reactor components are shown on the example of a pressurized water reactor and the requirements resulting therefrom for steel, the basic structural material, are derived. A detailed analysis of three main groups of reactor steels is presented and the applications are indicated of low-alloyed steels, high-alloyed austenitic steels, and steels with a high content of Ni and of alloying additions for steam generator pipes. An outline is given of prospective fast breeder reactor steels. (J.K.)

  8. Commercialization of fast reactors

    International Nuclear Information System (INIS)

    Comparative analysis has been performed of capital and fuel cycle costs for fast BN-type and pressurized light water VVER-type reactors. As a result of materials demand and components costs comparison of NPPs with VVER-1000 and BN-600 reactors, respectively, conclusion was made, that under equal conditions of the comparison, NPP with fast reactor had surpassed the specific capital cost of NPP with VVER by about 30 - 40 %. Ways were determined for further decrease of this difference, as well as for the fuel cycle cost reduction, because at present it is higher than that of VVER-type reactors. (author)

  9. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  10. Natural convection type reactor

    International Nuclear Information System (INIS)

    In a natural convection type nuclear reactor, a reactor core is disposed such that the top of the reactor core is always situated in a flooded position even if pipelines connected to the pressure vessel are ruptured and the level at the inside of the reactor vessel is reduced due to flashing. Further, a lower dry well situated below the pressure vessel is disposed such that it is in communication with a through hole to a pressure suppression chamber situated therearound and the reactor core is situated at the level lower than that of the through hole. If pipelines connected to the pressure vessel are ruptured to cause loss of water, although the water level is lowered after the end of the flashing, the reactor core is always flooded till the operation of a pressure accummulation water injection system to prevent the top of the reactor core even from temporary exposure. Further, injected water is discharged to the outside of the pressure vessel, transferred to the lower dry well, and flows through the through hole to the pressure control chamber and cools the surface of the reactor pressure vessel from the outside. Accordingly, the reactor core is cooled to surely and efficiently remove the after-heat. (N.H.)

  11. INVAP's Research Reactor Designs

    International Nuclear Information System (INIS)

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors IAEA safety

  12. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention provides a control device which can conduct scram and avoid lowering of the power of a nuclear power plant upon occurrence of earthquakes. Namely, the device of the present invention comprises, in addition to an existent power control device, (1) an earthquake detector for detecting occurrence and annihilation of earthquakes and (2) a reactor control device for outputting control rod operation signals and reactor core flow rate control signals depending on the earthquake detection signals from the detector, and reactor and plant information. With such a constitution, although the reactor is vibrated by earthquakes, the detector detects slight oscillations of the reactor by initial fine vibration waves as premonitory symptoms of serious earthquakes. The earthquake occurrence signals are outputted to the reactor control device. The reactor control device, receiving the signals, changes the position of control rods by way of control rod driving mechanisms to make the axial power distribution in the reactor core to a top peak type. As a result, even if the void amount in the reactor core is reduced by the subsequent actual earthquakes, since the void amount is moved, effects on the increase of neutron fluxes by the actual earthquakes is small. (I.S.)

  13. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  14. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  15. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  16. Latest results from the Cadarache 1 MV SINGAP experiment

    International Nuclear Information System (INIS)

    The SINGle Aperture - SINgle GAP (SINGAP) acceleration concept has been developed as a simplified alternative to the Multi-Aperture, Multi-Grid (MAMuG) acceleration of the ITER Neutral Beam reference design. The objective of the present experiments is to demonstrate reliable multi-second acceleration of a D- beam to 1 MeV (∼100 mA, 200 A/m2), relevant to the ITER neutral beam injection (NBI) requirements. During the previous studies it has been demonstrated that the SINGAP concept works and good quality 860 keV H- beams (43 mA, 1 s) and 630 keV D- beams (106 mA, 1 s) have been produced. In the present R and D programme another attempt has been made to reach the full beam energy and increase the extracted beam current density with a refurbished 1 MV epoxy insulator/bushing. Refurbishment had become necessary after the top two of the nine epoxy rings making up the insulator had been perforated and carbonised due to high voltage discharges. The best results achieved during the present campaign are 914 keV D- beams (58 mA, 1 s) with a D- current density of 30 A/m2 and 600 keV D- beams (76 mA, 1 s) with a D- current density of 70 A/m2. These results are world records. The beam optics is largely as expected from the modelling with the Vector Fields OPERA 3D code

  17. Installation of a shielded SIMS in CEA Cadarache

    International Nuclear Information System (INIS)

    A shielded SIMS 6f has been installed in the LECA in 1999 and will examine nuclear fuels before the end of this year. Meanwhile several studies have been realised to test its technical performance for nuclear materials analysis. After a brief presentation of the design of the shielded SIMS, five examples of preliminary results will be underlined. In the nuclear field, SIMS is currently used for the analysis of Zircaloy cladding corrosion because it can detect light elements like boron or lithium. More than a quantitative measurement of isotopes, it was possible with our SIMS to produce a two dimension image of the corrosion layer focusing on the protective layer. SIMS has also the advantage to detect isotopes with a low concentration. This property will be enlightened with some results obtained on sulphur in UO2. Because SIMS uses the erosion of the sample surface, it is possible to make depth profile. This possibility will be demonstrated on water altered glass samples on which the altered layer will be evidenced on depth profile. The last application of SIMS presented in the paper is addressed to nuclear fuel with detection of xenon. Fission gases release is indeed of first importance for nuclear fuel because it affects fuel swelling, rod internal pressure, fuel thermal conductivity via the formation of pores etc. We have demonstrated that it was possible to measure xenon on an unirradiated UO2 sample implanted with xenon. It was also proved that xenon could be measured in gaseous state, which would make possible the measurement of xenon filling pores. (author)

  18. Proceedings of the Cadarache Workshop on Electrostatic Turbulence

    International Nuclear Information System (INIS)

    Investigations in the field of confinement problems and anomalous transport in fusion devices are reported. Special attention is given to experimental and theoretical developments in the field of nonlinear physics and electrostatic turbulence. Results of experiments on scattering measurement techniques are also presented. Studies involving laser light scattering and microwave reflectometry are included

  19. One piece reactor removal

    International Nuclear Information System (INIS)

    Japan Research Reactor No.3 (JRR-3) was the first reactor consisting of 'Japanese-made' components alone except for fuel and heavy water. After reaching its initial critical state in September 1962, JRR-3 had been in operation for 21 years until March 1983. It was decided that the reactor be removed en-bloc in view of the work schedule, cost and management of the reactor following the removal. In the special method developed jointly by the Japanese Atomic Energy Research Institute and Shimizu Construction Co., Ltd., the reactor main unit was cut off from the building by continuous core boring, with its major components bound in the block with biological shield material (heavy concrete), and then conveyed and stored in a large waste store building constructed near the reactor building. Major work processes described in this report include the cutting off, lifting, horizontal conveyance and lowering of the reactor main unit. The removal of the JRR-3 reactor main unit was successfully carried out safely and quickly by the en-block removal method with radiation exposure dose of the workers being kept at a minimum. Thus the high performance of the en-bloc removal method was demonstrated and, in addition, valuable knowhow and other data were obtained from the work. (Nogami, K.)

  20. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  1. The fusion reactor

    International Nuclear Information System (INIS)

    Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

  2. Polymerization Reactor Engineering.

    Science.gov (United States)

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  3. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  4. Gas-cooled reactors

    International Nuclear Information System (INIS)

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  5. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  6. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  7. Light water type reactor

    International Nuclear Information System (INIS)

    The nuclear reactor of the present invention prevents disruption of a reactor core even in a case of occurrence of entire AC power loss event, and even if a reactor core disruption should occur, it prevents a rupture of the reactor container due to excess heating. That is, a high pressure water injection system and a low pressure water injection system operated by a diesel engine are disposed in the reactor building in addition to an emergency core cooling system. With such a constitution, even if an entire AC power loss event should occur, water can surely be injected to the reactor thereby enabling to prevent the rupture of the reactor core. Even if it should be ruptured, water can be sprayed to the reactor container by the low pressure water injection system. Further, if each of water injection pumps of the high pressure water injection system and the low pressure water injection system can be driven also by motors in addition to the diesel engine, the pump operation can be conducted more certainly and integrally. (I.S.)

  8. Naval propulsion reactors

    International Nuclear Information System (INIS)

    This article deals with the design and exploitation of naval propulsion reactors, mainly of PWR-type. The other existing or conceivable types of reactors are also presented: 1 - specificities of nuclear propulsion (integration in the ship, marine environment, maneuverability, instantaneous availability, conditions of exploitation-isolation, nuclear safety, safety authority); 2 - PWR-type reactor (stable operation, mastered technology, general design, radiation protection); 3 - other reactor types; 4 - compact or integrated loops architecture; 5 - radiation protection; 6 - reactor core; 7 - reactivity control (core lifetime, control means and mechanisms); 8 - core cooling (natural circulation, forced circulation, primary flow-rate program); 9 - primary loop; 10 - pressurizer; 11 - steam generators and water-steam secondary loop; 12 - auxiliary and safety loops; 13 - control instrumentation; 14 - operation; 15 - nuclear wastes and dismantling. (J.S.)

  9. Iris reactor conceptual design

    International Nuclear Information System (INIS)

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  10. Research reactor DHRUVA

    International Nuclear Information System (INIS)

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

  11. Reactor core monitoring method

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Michitsugu [Tokyo Electric Power Co., Inc. (Japan); Kanemoto, Shigeru; Enomoto, Mitsuhiro; Ebata, Shigeo

    1998-05-06

    The present invention provides a method of monitoring the state of coolant flow in a reactor of a BWR power plant. Namely, a plurality of local power region monitors (LPRM) are disposed to the inside of the reactor core for monitoring a power distribution. Signals of at least two optional LPRM detectors situated at positions different in axial or radial positions of the reactor core are obtained. General fluctuation components which nuclear hydrothermally fluctuate in overall reactor core are removed from the components of the signals. Then, correlational functions between these signals are determined. The state of coolant flow in the reactor is monitored based on the correlational function. When the axial flowing rate and radial flow interference are monitored, the accuracy upon monitoring axial and radial local behaviors of coolants can be improved by thus previously removing the general fluctuation components from signals of LPRM detectors and extracting local void information near to LPRM detectors at high accuracy. (I.S.)

  12. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  13. Mirror reactor surface study

    International Nuclear Information System (INIS)

    A general survey is presented of surface-related phenomena associated with the following mirror reactor elements: plasma first wall, ion sources, neutral beams, director converters, vacuum systems, and plasma diagnostics. A discussion of surface phenomena in possible abnormal reactor operation is included. Several studies which appear to merit immediate attention and which are essential to the development of mirror reactors are abstracted from the list of recommended areas for surface work. The appendix contains a discussion of the fundamentals of particle/surface interactions. The interactions surveyed are backscattering, thermal desorption, sputtering, diffusion, particle ranges in solids, and surface spectroscopic methods. A bibliography lists references in a number of categories pertinent to mirror reactors. Several complete published and unpublished reports on surface aspects of current mirror plasma experiments and reactor developments are also included

  14. FBR type reactor

    International Nuclear Information System (INIS)

    A circular neutron reflector is disposed vertically movably so as to surround the outer circumference of a reactor core barrel. A reflector driving device comprises a driving device main body attracted to the outer wall surface of the reactor barrel by electromagnetic attraction force and an inertia body disposed above the driving device main body vertically movably. A reflector is connected below the reactor driving device. At the initial stage, a spontaneous large current is supplied to upper electromagnetic repulsion coils of the reflector driving device, impact electromagnetic repulsion force is caused between the inertia body and the reflector driving device, so that the driving device main body moves downwardly by a predetermined distance and stopped. The reflector driving device can be lowered in a step-like manner to an appropriate position suitable to restart the reactor during stoppage of the reactor core by conducting spontaneous supply of current repeatedly to the upper electromagnetic repulsion coils. (I.N.)

  15. TRIGA research reactors

    International Nuclear Information System (INIS)

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  16. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  17. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  18. Compact torsatron reactors

    International Nuclear Information System (INIS)

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

  19. Multi-purpose reactor

    International Nuclear Information System (INIS)

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MWth, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co60) production capacity is 50000 Ci/yr, 200Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  20. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  1. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  2. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  3. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  4. First ITER Council convened in Cadarache - historic step in the quest for clean Energy 28 November 2007 Cadarache

    CERN Multimedia

    Hay, Jennifer

    2007-01-01

    "On 27 November 2007, the ITER Council convened for the first time in the history of the new International Organization. Opening the meeting, Dr Werner Burkart, Deputy Director of the International Atomic Energy Agency (AIEA), said: "Let me congratulate all who have contributed to the achievements of the ITER initiative to date.... (1,5 page)

  5. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  6. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  7. Safety of research reactors

    International Nuclear Information System (INIS)

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

  8. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  9. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  10. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-06-01

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

  11. FBR type reactor

    International Nuclear Information System (INIS)

    The present invention provides an FBR type reactor in which the combustion of reactor core fuels is controlled by reflectors, and the position of a reflector driving device can be controlled even during shut down of the reactor. Namely, the reflector driving device is attracted to the outer wall surface of a reactor core barrel by electromagnetic attraction force. An inertia body is disposed vertically movably to the upper portion of the reflector driving device. Magnetic repulsive coils generate instantaneous magnetic repulsive force between the inertia body and the reflector driving device. With such a constitution, the reflector driving device can be driven by using magnetic repulsion of the electromagnetic repulsive coils and inertia of the inertia body. As a result, not only the reflectors can be elevated at an ultraslow speed during normal reactor operation, but also fine position adjustment for the reflector driving device, as well as fine position adjustment of the reflectors required upon restart of the reactor can be conducted by lowering the reflector driving device during shut down of the reactor. (I.S.)

  12. Reactor water sampling device

    International Nuclear Information System (INIS)

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  13. Test reactor technology

    International Nuclear Information System (INIS)

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

  14. Advanced reactor licensing issues

    International Nuclear Information System (INIS)

    In July 1986 the US Nuclear Regulatory Commission issued a Policy Statement on the Regulation of Advanced Nuclear Power Plants. As part of this policy advanced reactor designers were encouraged to interact with NRC early in the design process to obtain feedback regarding licensing requirements for advanced reactors. Accordingly, the staff has been interacting with the Department of Energy (DOE) and its contractors on the review of three advanced reactor conceptual designs: one modular High Temperature Gas-Cooled Reactor (MHTGR) and two Liquid Metal Reactors (LMRs). As a result of these interactions certain safety issues associated with these advanced reactor designs have been identified as key to the licensability of the designs as proposed by DOE. The major issues in this regard are: (1) selection and treatment of accident scenarios; (2) selection of siting source term; (3) performance and reliability of reactor shutdown and decay heat removal systems; (4) need for conventional containment; (5) need for conventional emergency evacuation; (6) role of the operator; (7) treatment of balance of plant; and (8) modular approach. This paper provides a status of the NRC review effort, describes the above issues in more detail and provides the current status and approach to the development of licensing guidance on each

  15. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  16. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  17. Determination of research reactor safety parameters by reactor calculations

    International Nuclear Information System (INIS)

    Main research reactor safety parameters such as power density peaking factors, shutdown margin and temperature reactivity coefficients are treated. Reactor physics explanation of the parameters is given together with their application in safety evaluation performed as part of research reactor operation. Reactor calculations are presented as a method for their determination assuming use of widely available computer codes. (author)

  18. Reactor de plasma

    OpenAIRE

    Erra Serrabasa, Pilar; Molina Mansilla, Ricardo; Beltrán Serra, Eric

    2008-01-01

    Reactor de plasma. Se trata de un reactor de plasma que puede trabajar en un amplio rango de presión, desde el vacío y presiones reducidas hasta la presión atmosférica y presiones superiores. Adicionalmente el reactor de plasma tiene la capacidad de regular otros parámetros importantes y permite su uso para el tratamiento de muestras de tipología muy diversa, como por ejemplo las de tamaño relativamente grande o de superficie rugosa.

  19. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  20. Microfluidic electrochemical reactors

    Science.gov (United States)

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  1. Licensed operating reactors

    International Nuclear Information System (INIS)

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  2. First Algerian research reactor

    International Nuclear Information System (INIS)

    In 1985, both the Algerian Commissariat of New Energies and the Argentine National Atomic Energy Commission plus the firm INVAP S.E., started a series of mutual visits aimed at defining the mechanisms for cooperation in the nuclear field. Within this framework, a commercial contract was undersigned covering the supply of a low-power reactor (RUN), designed for basic and applied research in the fields of reactor physics and nuclear engineering. The reactor may also be used for performing experiences with neutron beams, for the irradiation of several materials and for the training of technicians, scientists and operators

  3. Course on reactor physics

    International Nuclear Information System (INIS)

    In Germany only few students graduate in nuclear technology, therefore the NPP operating companies are forced to develop their own education and training concepts. AREVA NP has started together with the Technical University of Dresden a one-week course ''reactor physics'' that includes the know-how of the nuclear power plant construction company. The Technical University of Dresden has the training reactor AKR-2 that is retrofitted by modern digital instrumentation and control technology that allows the practical training of reactor control.

  4. Fast Breeder Reactor studies

    International Nuclear Information System (INIS)

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts

  5. Nuclear reactor theory

    International Nuclear Information System (INIS)

    This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

  6. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  7. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

  8. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  9. Joyo experimental reactor tour

    International Nuclear Information System (INIS)

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

  10. Fast Breeder Reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  11. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  12. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  13. Reactor monitoring system

    International Nuclear Information System (INIS)

    The present invention concerns a device for monitoring the inside of an FBR type reactor which can not be monitored by a usual optical camera. An ultrasonic camera having an excellent propagating property in a liquid metal sodium is scanned, and reflected waves of the ultrasonic waves are received as signals. The signals are processed by using a virtual realistic feeling (VR) technique such as a head mounting type image display (HMD) and a three dimensional pointing device. With such procedures, the inside of the FBR type reactor can be observed with such a realistic feeling that the inside of the FBR type reactor were seen directly. (I.S.)

  14. Research reactor support

    International Nuclear Information System (INIS)

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  15. Study of power reactor dynamics by stochastic reactor oscillator method

    International Nuclear Information System (INIS)

    Stochastic reactor oscillator and cross correlation method were used for determining reactor dynamics characteristics. Experimental equipment, fast reactor oscillator (BOR-1) was activated by random pulses from the GBS-16 generator. Tape recorder AMPEX-SF-300 and data acquisition tool registered reactor response to perturbations having different frequencies. Reactor response and activation signals were cross correlated by digital computer for different positions of stochastic oscillator and ionization chamber

  16. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  17. Reactor power control device

    International Nuclear Information System (INIS)

    The present invention concerns a method of controlling reactor power to shift it into a partial power operation upon occurrence of recycling pump tripping or loss of generator load. Operation state of a reactor is classified into a plurality of operation states based on values of the reactor core flow rate and the reactor power. Different insertion patterns for selected control rods are determined on every classified operation states. Then, an insertion pattern corresponding to the operation state upon occurrence of recycling pump tripping or loss of power generator load is carried out to shift into partial power operation. The operation is shifted to a load operation solely in the station while avoiding risks such as TPM scram. Then neutron fluxes are suppressed upon transient to increase margin of fuel integrity. Selected control rod pattern of the optimum reactivity is set to each of operation regions, thereby enabling to conduct flexible countermeasure so as to attain optimum operationability. (N.H.)

  18. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  19. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2001-04-01

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

  20. Reactor BR2. Introduction

    Energy Technology Data Exchange (ETDEWEB)

    Gubel, P

    2002-04-01

    The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

  1. Reactor parameter simulation system

    International Nuclear Information System (INIS)

    A reactor parameter simulation system (RPSS) has been built with the capability of analyzing any reactor signals, decomposing those signals into their deterministic and stochastic components, then reconstructing new, simulated signals that possess the same statistical and correlation structure as the original plant variables. Important uses of the RPSS are for integration with reactor simulation software to provide tools for plant control strategy development, and for safety-study investigations of scenarios that can arise involving signal faults generated from degraded sensors. A third use of the RPSS is for frequency-domain filtering of reactor process variables contaminated with serially correlated noise, which is important for our ongoing development of expert systems for sensor-operability surveillance. 5 refs., 4 figs., 3 tabs

  2. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  3. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  4. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  5. Ageing of research reactors

    International Nuclear Information System (INIS)

    Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)

  6. Experience with Kamini reactor

    International Nuclear Information System (INIS)

    Kamini is a 233U fuelled, 30 kW(th) research reactor. It is one of the best neutron source facility with a core average flux of 1012 n/cm2/s in IGCAR used for neutron radiography of active and nonradioactive objects, activation analysis and radiation physics research. The core consists of nine plate type fuel elements with a total fuel inventory of 590 g of 233U. Two safety control plates made of cadmium are used for start up and shutdown of the reactor. Three beam tubes, two-thimble irradiation site outside reflector and one irradiation site nearer to the core constitute the testing facilities of Kamini. Kamini attained first criticality on 29th October 96 and nominal power of 30 kW in September 1997. This paper covers the design features of the reactor, irradiation facilities and their utilities and operating experience of the reactor. (author)

  7. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  8. Future Reactor Experiments

    CERN Document Server

    He, Miao

    2013-01-01

    The measurement of the neutrino mixing angle $\\theta_{13}$ opens a gateway for the next generation experiments to measure the neutrino mass hierarchy and the leptonic CP-violating phase. Future reactor experiments will focus on mass hierarchy determination and the precision measurement of mixing parameters. Mass hierarchy can be determined from the disappearance of reactor electron antineutrinos based on the interference effect of two separated oscillation modes. Relative and absolute measurement techniques have been explored. A proposed experiment JUNO, with a 20 kton liquid scintillator detector of $3%/$$\\sqrt{E(MeV)}$ energy resolution, $\\sim$ 53 km far from reactors of $\\sim$ 36 GW total thermal power, can reach to a sensitivity of $\\Delta\\chi^{2}>16$ considering the spread of reactor cores and uncertainties of the detector response. Three of mixing parameters are expected to be measured to better than 1% precision. There are multiple detector options for JUNO under investigation. The technical challenges...

  9. Reactor hot spot analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  10. Reactor BR2. Introduction

    International Nuclear Information System (INIS)

    The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

  11. Research Reactor Benchmarks

    International Nuclear Information System (INIS)

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given

  12. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  13. Reactor Neutrino Spectra

    CERN Document Server

    Hayes, A C

    2016-01-01

    We present a review of the antineutrino spectra emitted from reactors. Knowledge of these and their associated uncertainties are crucial for neutrino oscillation studies. The spectra used to-date have been determined by either conversion of measured electron spectra to antineutrino spectra or by summing over all of the thousands of transitions that makeup the spectra using modern databases as input. The uncertainties in the subdominant corrections to beta-decay plague both methods, and we provide estimates of these uncertainties. Improving on current knowledge of the antineutrino spectra from reactors will require new experiments. Such experiments would also address the so-called reactor neutrino anomaly and the possible origin of the shoulder observed in the antineutrino spectra measured in recent high-statistics reactor neutrino experiments.

  14. Pulsed fusion reactors

    International Nuclear Information System (INIS)

    This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

  15. Reactor fueling of BWR type reactors

    International Nuclear Information System (INIS)

    Purpose: To enable the pattern exchange for control rods during burning in Control Cell Core type BWR reactors. Constitution: A plurality of control cells are divided into a plurality of groups such that the control cells is aparted from each other by way of at least two fuel assemblies other than the control cells with respect to the vertical and lateral directions of the reactor core cross section, as well as they are in adjacent with control cells of other groups with respect to the orthogonal direction. This enables to perform the pattern exchange for the control rods during burning in the control cell core with ease, and the control blade and the story effect harmful to the mechanical soundness of fuels can thus be suppressed. (Moriyama, K.)

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  17. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  18. The replacement research reactor

    International Nuclear Information System (INIS)

    The contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000. This was followed by the completion of the detailed design and an application for a construction licence was made in May 2001. This paper will describe the main elements of the design and their relation to the proposed applications of the reactor. The future stages in the project leading to full operation are also described

  19. OECD Halden reactor project

    International Nuclear Information System (INIS)

    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

  20. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  1. Small reactor return

    International Nuclear Information System (INIS)

    Current state of the development of present-day small reactors in different countries is performed. Various designs of low and middle power reactors, among which are CAREM (25 MW, PWR), KLT-40 (40 MW, PWR), MRX (30 MW, PWR), IRIS (50 MW, PWR), SMART (1000 MW, PWR), Modular SBWR (50 MW, BWR), PBMR (120 MW, HTGR), GT-HMR (285 MW, HTGR), are discussed

  2. Reactor lattice transport calculations

    International Nuclear Information System (INIS)

    The present lecture is a continuation of the lecture on Introduction to the Neutron Transport Phenomena. It comprises three aspects of lattice calculations. First the idea of a reactor lattice is introduced. Then the main definitions used in reactor lattice analysis are given, and finally two basic methods applied for solution of the transport equations are defined. Several remarks on secondary results from lattice transport calculations are added. (author)

  3. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  4. Future reactor experiments

    International Nuclear Information System (INIS)

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper

  5. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  6. Jet-Stirred Reactors

    OpenAIRE

    Herbinet, Olivier; Guillaume, Dayma

    2013-01-01

    The jet-stirred reactor is a type of ideal continuously stirred-tank reactor which is well suited for gas phase kinetic studies. It is mainly used to study the oxidation and the pyrolysis of hydrocarbon and oxygenated fuels. These studies consist in recording the evolution of the conversion of the reactants and of the mole fractions of reaction products as a function of different parameters such as reaction temperature, residence time, pressure and composition of the inlet gas. Gas chromatogr...

  7. Generation IV reactors: economics

    International Nuclear Information System (INIS)

    The operating nuclear reactors were built over a short period: no more than 10 years and today their average age rounds 18 years. EDF (French electricity company) plans to renew its reactor park over a far longer period : 30 years from 2020 to 2050. According to EDF this objective implies 3 constraints: 1) a service life of 50 to 60 years for a significant part of the present operating reactors, 2) to be ready to built a generation 3+ unit in 2020 which infers the third constraint: 3) to launch the construction of an EPR (European pressurized reactor) prototype as soon as possible in order to have it operating in 2010. In this scheme, generation 4 reactor will benefit the feedback experience of generation 3 and will take over in 2030. Economic analysis is an important tool that has been used by the generation 4 international forum to select the likely future reactor systems. This analysis is based on 4 independent criteria: the basic construction cost, the construction time, the operation and maintenance costs and the fuel cycle cost. This analysis leads to the evaluation of the global cost of electricity generation and of the total investment required for each of the reactor system. The former defines the economic competitiveness in a de-regulated energy market while the latter is linked to the financial risk taken by the investor. It appears, within the limits of the assumptions and models used, that generation 4 reactors will be characterized by a better competitiveness and an equivalent financial risk when compared with the previous generation. (A.C.)

  8. Future reactor experiments

    Science.gov (United States)

    Wen, Liangjian

    2015-07-01

    The non-zero neutrino mixing angle θ13 has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  9. Department of Reactor Technology

    DEFF Research Database (Denmark)

    Risø National Laboratory, Roskilde

    The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included.......The general development of the Department of Reactor Technology at Risø during 1981 is presented, and the activities within the major subject fields are described in some detail. Lists of staff, publications, and computer programs are included....

  10. AVR reactor physics

    International Nuclear Information System (INIS)

    A process for reactivity control was developed and used for fuelling the AVR reactor core, which is largely based on experimentally determined values. By adding fuel elements with different quantities of heavy metals paired with various experimental requirements, great demands were made of reactivity control. Although only a small range of control was available, this was sufficient to operate the reactor and to shut it down safely in the required power and temperature range. (orig.)

  11. Moon base reactor system

    Science.gov (United States)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  12. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  13. Emergency reactor scram system

    International Nuclear Information System (INIS)

    The present invention provides an emergency reactor scram system capable of shut down a reactor safely upon occurrence of pump trip by improving a passive scram performance for an FBR-type reactor. Namely, a driving motor and an electric generator are connected to a main pump of a primary system. An AC/DC convertor is connected to the electric generator. A shielding plug is disposed to the upper end opening of a reactor container, a control rod drive mechanism is erected on the shielding plug, and an extension pipe is attached to scram magnets of the control rod drive mechanism. The extension pipe is connected to a control rod. The rotation of the shaft of the pump is used as a direct rotator to provide an integrated-type electric generator. The electric generator is electrically connected with the power source of a scram magnet of the emergency scram system. Accordingly, the control rod of the emergency scram system is automatically and rapidly inserted to the reactor core using the power source of the electric generator upon trip of the main pump thereby enabling to scram the reactor safely. (I.S.)

  14. A modular reactor plant

    International Nuclear Information System (INIS)

    This paper describes a new concept in liquid metal reactors that is being developed by General Electric under contract to the Department of Energy. This concept is called the Modular Reactor Plant. While this effort is not expected to have a near-term impact, it is directed toward three principal issues currently affecting nuclear power in the United States. First, plant costs have escalated to the point where the startup of new plants require large electric rate increases. Second, the cost of new plants coming on-line today vary by as much as a factor of three. And, third, nuclear construction times often exceed the utilities prudent planning cycle. This paper describes how General Electric's Modular Reactor Plant addreses these issues through shop fabrication and assembly, rail shipment to the site for rapid installation of nuclear components and inherent reactor protection. In addition, it is expected the modular reactor plant will reduce the current cost of development and demonstration of liquid metal reactors to an affordable level

  15. New fission reactor designs

    International Nuclear Information System (INIS)

    A number of critical challenges to the expanded or continued use of nuclear power have developed. These can be categorized as: regulatory restrictions and complications; negative public attitudes; plant complexity; plant life, operations, and maintenance; uncertain load growth, financing; waste management. Solutions to these challenges through advanced reactor design centre around four key technical responses. Passive safety systems are being introduced which use the laws of physics to provide emergency reactor coding, control and shutdown thus eliminating the possibility of human error. Modular construction promises cuts in costs and construction time by shifting the major part of component manufacture from the site to the factory. Standardization also cuts capital costs and in addition operations and repair costs and expedites reactor licensing. Improvements to the fuel cycle include improved fuel types, designs and fabrication, and the reprocessing of and recycling spent fuel back into energy production, thus extending uranium resources and offering a partial solution to the problem of waste disposal. Examples of evolutionary and advanced water-cooled reactors, modular high temperature gas-cooled reactors, and advanced liquid metal cooled fast breeder reactors which are being developed round the world are presented. (author)

  16. OECD Halden reactor project

    International Nuclear Information System (INIS)

    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

  17. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention efficiently calibrates a fixed type gamma ray thermometer of a reactor power measuring device of a BWR type reactor. Namely, the device of the present invention calculates peripheral fuel rod power distribution by calibrating the reactor power distribution by heat generation amount, the reactor power distribution being obtained by a calculation based on a reactor model for converting the signals of a plurality of the gamma ray thermometers in the reactor core based on a conversion formula. In this case, the conversion formula is a relational formula between the power of a thermocouple of the gamma ray thermometer, gamma ray heat generation amount, thermocouple zero power sensitivity relative to a temperature coefficient. A conversion efficient calculation means makes a calibration heater to generate heat at a predetermined power, and the thermocouple zero power sensitivity and the temperature coefficient are obtained based on the output of the gamma ray thermometer in this case. The calibration means updates to conversion type thermocouple zero power sensitivity and temperature coefficient. A calibration execution means executes the operations described above successively, and when the thermocouple zero power sensitivity and the temperature coefficient are out of an allowable range, the means informs it and eliminates the corresponding gamma ray thermometer from the measuring meters. (I.S.)

  18. Reactor safety engineering

    International Nuclear Information System (INIS)

    The concept of the work is such that the basic safety philosophy for nuclear power plants as well as the safety features of both types of light water reactors, pressurized and boiling water reactors, and of the fast breeder reactor are dealt with. With the pressurized and boiling water reactors also variations, due to different supplies are mentioned. The state of development considered is characterized by the results of the American reactor safety study having very much influenced the way of presentation and the validity of the information contained. In the introduction the attentive reader is made familiar with the basic traits of safety engineering, the traditional deterministic way of proceeding being supplemented by a detailed illustration of probabilistic means used in the safety analysis. Added to this are comparative descriptions of the individual safety features, their design and mode of operation. There are, e.g., detailed discussion of the emergency core cooling systems, the power supply systems, the reactor protection system, and the containment. Special chapters are attributed to transients with and without the fast shutdown system working and to loss of coolant. The so-called external events are treated somewhat shortly whereas much space is given to core melting problems. The treatment of important events from the safety point of view, including the section on Harrisburg added for reasons of immediate interest, is limited to phenomenological description. (orig.)

  19. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

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

    International Nuclear Information System (INIS)

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

  1. REACTOR GROUT THERMAL PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  2. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  3. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  4. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  5. JEFF-3.1.1 nuclear data validation for sodium fast reactors

    International Nuclear Information System (INIS)

    The JEFF-3.1.1 Nuclear Data Library is the latest version of the Joint Evaluated Fission and Fusion Library. The complete suite of data was released in 2008, and contains general purpose nuclear data evaluations compiled at the NEA Data Bank in co-operation with several laboratories in NEA Data Bank member countries. JEFF-3.1.1 contains also radioactive decay data, activation data and fission yields data. It combines the efforts of the JEFF and EFF Working Groups who have contributed to this combined fission and fusion file. The library contains neutron reaction data, incident proton data and thermal neutron scattering law data in the ENDF-6 format. The aim of this paper is to present the status of the validation of this library using the Monte Carlo Code TRIPOLI4.5 for fast reactor calculations. To reach that goal, we reanalyse a selected set of integral experiments performed in MASURCA Mock-up at CEA/CADARACHE, in ZPPR mock-up at INL USA and in SUPERPHENIX power Reactor. These experiments are: - The CIRANO program in MASURCA (1994-1997) was meant to extend the validation of ERANOS (code, schemes, data libraries) to Pu-burning fast reactors (CAPRA project) via the progressive substitution of fertile blankets by steel reflectors; - The ZPPR10A experiment proposed in IRPhE of the NEA Data Bank. This experiment is complementary of the first one because sodium void effects have been measured and are available; - Several experiments made during de commissioning of SUPERPHENIX that give information on different types of critical states of the core. All these experiments are modelled with the TRIPOLI code to avoid most of the errors due to deterministic models and to focus only on the nuclear data biases. An example of the SUPERPHENIX core modelling is given on the figure 1. Ongoing analysis shows the capability of the new JEFF3.1.1 nuclear library to predict the SFR neutronic behaviours.From this work and from the qualification work performed with ERANOS2, some required

  6. Lead-cooled hybrid reactors and fuel regeneration for energy production and incineration evolution of physical parameters and induced radiotoxicity; Capacites des reacteurs hybrides au plomb pour la production d'energie et l'incineration avec multirecyclage des combustibles evolution des parametres physiques radiotoxicites induites

    Energy Technology Data Exchange (ETDEWEB)

    David, S

    1999-07-01

    The concept of accelerator driven subcritical reactors (hybrid reactors), as re-launched in the beginning of the 1990's by C. Rubbia and C.D. Bowman, allows to open new paths in the management of radioactive wastes. This work treats, first, of the study of the neutron multiplication characteristics in a subcritical reactor core and shows the fundamental differences with critical systems and the advantages that follow. This study is based on the series of measurements performed at Cadarache (Muse experiment), the first results of which are presented. The subcritical property of an hybrid reactor makes this system very flexible and allows to foresee different uses, like the energy production or the incineration of wastes. The second part of this work deals with the Monte Carlo simulation of the capacities of fast spectrum and lead-cooled hybrid systems to produce energy by using different fuel cycles (uranium and thorium), and in the same time regenerating the fissile matter and keeping the reactivity up without any external intervention. Different types of fuel multi-recycles are considered. The results allow to quantify the advantages linked with the use of the thorium cycle, in particular in terms of radiotoxicity abatement. The study of the intermediate steps necessary to develop this reactor technology with the present day fuels (plutonium from thermal reactors and enriched uranium) proposes an efficient management of the actinides produced by today's reactors which are used as auxiliary fissile materials. Finally, the incineration of actinides at the end of the cycle (shutdown scenario) is considered and allows to describe the advantage of lead-cooled hybrid systems for the abatement of the radiotoxicity of an inventory at the end of cycle. (J.S.)

  7. BR2 Reactor: Introduction

    International Nuclear Information System (INIS)

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

  8. Reactor coolant cleanup facility

    International Nuclear Information System (INIS)

    A depressurization device is disposed in pipelines upstream of recycling pumps of a reactor coolant cleanup facility to reduce a pressure between the pressurization device and the recycling pump at the downstream, thereby enabling high pressure coolant injection from other systems by way of the recycling pumps. Upon emergency, the recycling pumps of the coolant cleanup facility can be used in common to an emergency reactor core cooling facility and a reactor shutdown facility. Since existent pumps of the emergency reactor core cooling facility and the reactor shutdown facility which are usually in a stand-by state can be removed, operation confirmation test and maintenance for equipments in both of facilities can be saved, so that maintenance and reliability of the plant are improved and burdens on operators can also be mitigated. Moreover, low pressure design can be adopted for a non-regenerative heat exchanger and recycling coolant pumps, which enables to improve the reliability and economical property due to reduction of possibility of leakage. (N.H.)

  9. HTGR type reactor

    International Nuclear Information System (INIS)

    A reactor core is disposed at the center of a reactor container, a reflector is disposed on the outer side thereof, a steam generator is disposed further outer side thereof coaxially, and they are constituted as an integrated one container. A gas circulator and control rod drives are protruded at the outer side of the lower portion of the integrated container. Heat insulators are disposed on the inner side of the container wall in the upper portion of the reactor container. Helium gas risen in the reactor core and heated to a high temperature descends in a circular steam generator and undergoes heat exchange with water, and is then pressurized in the gas circulator after the lowering of the temperature, and returned to the inlet of the reactor core from the lower central portion of the container. With such procedures, the helium gas as primary coolants circulates only in the container to improve confinement. The device can be reduced in the size and the cost. (I.N.)

  10. Reactor container spray device

    International Nuclear Information System (INIS)

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  11. PROTEUS research reactor

    International Nuclear Information System (INIS)

    The PROTEUS zero power reactor at the Paul Scherrer Institute (PSI) in Switzerland achieved first criticality in 1968 and since then has been operated as an experimental tool for reactor physics research on test lattices representative of a wide range of reactor concepts. Reactor design codes and their associated data libraries are validated on the basis of the experimental results obtained. PROTEUS is normally configured as a driven system, in which a subcritical test zone is made critical by the surrounding driver zones. The advantages of driven systems can be summarized as follows: - Smaller amount of test fuel is required; - Large range of test zone conditions (including k∞ < 1 states) can be investigated by changes in the driver loading alone, thus avoiding undesirable perturbations to the test zone which would influence the measurement conditions and thus affect the interpretability of the results; - Necessary reactor control and instrumentation equipment (usually perturbing from the experimental viewpoint) can be located in the outer driver regions, thereby avoiding disturbance of the test lattice

  12. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

  13. China experimental fast reactor

    International Nuclear Information System (INIS)

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

  14. EBT reactor analysis

    International Nuclear Information System (INIS)

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this operating window, physics and engineering systems analysis and cost sensitivity studies indicate that reactors with approx. 6 to 10%, P approx. 1200 to 1700 MW(e), wall loading approx. 1.0 to 2.5 MW/m2, and recirculating power fraction (including ring-sustaining power and all other reactors auxiliaries) approx. 10 to 15% are possible. A number of concept improvements are also proposed that are found to offer the potential for further improvement of the reactor size and parameters. These include, but are not limited to, the use of: (1) supplementary coils or noncircular mirror coils to improve magnetic geometry and reduce size, (2) energetic ion rings to improve ring power requirements, (3) positive potential to enhance confinement and reduce size, and (4) profile control to improve stability and overall fusion power density

  15. First experiments results about the engineering model of Rapsodie; Premiers resultats d'essais interessant le bloc pile de rapsodie

    Energy Technology Data Exchange (ETDEWEB)

    Chalot, A.; Ginier, R.; Sauvage, M. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report deals with the first series of experiments carried out on the engineering model of Rapsodie and on an associated sodium facility set in a laboratory hall of Cadarache. It conveys more precisely: 1/ - The difficulties encountered during the erection and assembly of the engineering model and a compilation of the results of the first series of experiments and tests carried out on this installation (loading of the subassemblies preheating, thermal chocks...). 2/ - The experiments and tests carried out on the two prototypes control rod drive mechanisms which brought to the choice for the design of the definitive drive mechanism. As a whole, the results proved the validity of the general design principles adopted for Rapsodie. (authors) [French] Ce rapport traite des premiers essais realises sur la maquette du bloc pile de Rapsodie et sur une installation annexe de sodium, implantees dans un hall d'essais de Cadarache. Il fait part: 1/- Des difficultes eprouvees lors du montage de la maquette et rassemble les resultats des premiers essais effectues sur cette installation (chargement des assemblages, prechauffage, chocs thermiques...). 2/- Des essais realises sur deux prototypes de mecanisme de barre de controle qui ont conduit a la conception du mecanisme definitif. L'ensemble des resultats obtenus a permis de confirmer la validite des principes adoptes pour la pile Rapsodie. (auteurs)

  16. Modern research reactors in the world and RA research reactor

    International Nuclear Information System (INIS)

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  17. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  18. Scaleable, High Efficiency Microchannel Sabatier Reactor Project

    Data.gov (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  19. Methanation assembly using multiple reactors

    Science.gov (United States)

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  20. Mimic of OSU research reactor

    International Nuclear Information System (INIS)

    The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor

  1. MINT research reactor safety program

    Energy Technology Data Exchange (ETDEWEB)

    Mohamad Idris bin Taib [Division of Special Project, Malaysian Institute for Nuclear Technology Research (MINT), Bangi (Malaysia)

    2000-11-01

    Malaysian Institute for Nuclear Technology Research (MINT) Research Reactor Safety Program has been done along with Reactor Power Upgrading Project, Reactor Safety Upgrading Project and Development of Expert System for On-Line Nuclear Process Control Project. From 1993 up to date, Neutronic and Thermal-hydraulics analysis, Probabilistic Safety Assessment as well as installation of New 2 MW Secondary Cooling System were done. Installations of New Reactor Building Ventilation System, Reactor Monitoring System, Updating of Safety Analysis Report and Upgrading Primary Cooling System are in progress. For future activities, Reactor Modeling will be included to add present activities. (author)

  2. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  3. Fusion reactor materials

    International Nuclear Information System (INIS)

    This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

  4. Chernobyl reactor accident

    International Nuclear Information System (INIS)

    On April 26, 1986, an explosion occurred at the newest of four operating nuclear reactors at the Chernobyl site in the USSR. The accident initiated an international technical exchange of almost unprecedented magnitude; this exchange was climaxed with a meeting at the International Atomic Energy Agency in Vienna during the week of August 25, 1986. The meeting was attended by more than 540 official representatives from 51 countries and 20 international organizations. Information gleaned from that technical exchange is presented in this report. A description of the Chernobyl reactor, which differs significantly from commercial US reactors, is presented, the accident scenario advanced by the Russian delegation is discussed, and observations that have been made concerning fission product release are described

  5. International Thermonuclear Experimental Reactor

    International Nuclear Information System (INIS)

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

  6. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  7. Licensed operating reactors

    International Nuclear Information System (INIS)

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  8. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  9. Colliding Beam Fusion Reactors

    Science.gov (United States)

    Rostoker, Norman; Qerushi, Artan; Binderbauer, Michl

    2003-06-01

    The recirculating power for virtually all types of fusion reactors has previously been calculated [1] with the Fokker-Planck equation. The reactors involve non-Maxwellian plasmas. The calculations are generic in that they do not relate to specific confinement devices. In all cases except for a Tokamak with D-T fuel the recirculating power was found to exceed the fusion power by a large factor. In this paper we criticize the generality claimed for this calculation. The ratio of circulating power to fusion power is calculated for the Colliding Beam Reactor with fuels D-T, D-He3 and p-B11. The results are respectively, 0.070, 0.141 and 0.493.

  10. The MNSR reactor

    International Nuclear Information System (INIS)

    This tank-in-pool reactor is based on the same design concept as the Canadian Slowpoke. The core is a right circular cylinder, 24 cm diameter by 25 cm long, containing 411 fuel pin positions. The pins are HEU-Aluminium alloy, 0.5 cm in diameter. Critical mass is about 900 g. The reactor has a single cadmium control rod. The back-up shutdown system is the insertion of a cadmium capsule in a core position. Excess reactivity is limited to 3.5mk. In both the MNSR and Slowpoke, the insertion of the maximum excess reactivity results in a power transient limited by the coolant/moderator temperature to safe values, independent of any operator action. This reactor is used primarily in training and neutron activation analysis. Up to 64 elements have been analyzed in a great variety of different disciplines. (author)

  11. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  12. Reactor operation experience

    International Nuclear Information System (INIS)

    Since the TRIGA Users Conference in Helsinki 1970 the TRIGA reactor Vienna was in operation without any larger undesired shutdown. The integrated thermal power production by August 15 1972 accumulated to 110 MWd. The TRIGA reactor is manly used for training of students, for scientific courses and research work. Cooperation with industry increased in the last two years either in form of research or in performing training courses. Close cooperation is also maintained with the IAEA, samples are irradiated and courses on various fields are arranged. Maintenance work was performed on the heat exchanger and to replace the shim rod magnet. With the view on the future power upgrading nine fuel elements type 110 have been ordered recently. Experiments, performed currently on the reactor are presented in details

  13. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    The development of a nuclear rocket engine reactor (NRER ) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

  14. Reactor accidents in perspective

    International Nuclear Information System (INIS)

    In each of the three major reactor accidents which have led to significant releases to the environment, and discussed in outline in this note, the reactor has been essentially destroyed - certainly Windscale and Chernobyl reactors will never operate and the cleanup operation for Three Mile Island is currently estimated to have cost in excess of US Pound 500 000 000. In each of the accidents there has not been any fatality off site in the short term and any long-term health detriment is unlikely to be seen in comparison with the natural cancer incidence rate. At Chernobyl, early fatalities did occur amongst those concerned with fighting the incident on site and late effects are to be expected. The assumption of a linear non-threshold risk, and hence no level of zero risk is the main problem in communication with the public, and the author calls for simplification of the presentation of the concepts of radiological protection. (U.K.)

  15. Reactor safety equipments

    International Nuclear Information System (INIS)

    Purpose: To positively recover radioactive substances discharged in a dry well at the time of failure of a reactor. Constitution: In addition to the emergency gas treating system fitted to a reactor building, a purification system connected through a pipeline to the dry well is arranged in the reactor building. This purification system is connected through pipes fitted to the dry well to forced circulation device, heat exchanger, and purification device. The atmosphere of high pressure steam gases in the dry well is derived to the heat exchanger for cooling, and then radioactive substances which are contained in the gases are removed by filter sets charged with the HEPA filters and the HECA filters. At last, there gases are returned to dry well by circulation pump, repeat this process. (Kamimura, M.)

  16. Licensed operating reactors

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  17. Reactor protection system

    International Nuclear Information System (INIS)

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  18. Backfitting swimming pool reactors

    International Nuclear Information System (INIS)

    Calculations based on measurements in a critical assembly, and experiments to disclose fuel element surface temperatures in case of accidents like stopping of primary coolant flow during full power operation, have shown that the power of the swimming pool type research reactor FRG-2 (15 MW, operating since 1967) might be raised to 21 MW within the present rules of science and technology, without major alterations of the pool buildings and the cooling systems. A backfitting program is carried through to adjust the reactor control systems of FRG-2 and FRG-1 (5 MW, housed in the same reactor hall) to the present safety rules and recommendations, to ensure FRG-2 operation at 21 MW for the next decade. (author)

  19. MERCHANT MARINE SHIP REACTOR

    Science.gov (United States)

    Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

    1961-05-01

    A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

  20. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  1. Study of future reactors

    International Nuclear Information System (INIS)

    Today, more than 420 large reactors with a gross output of close to 350 GWe supply 20 percent of world electricity needs, accounting for less than 5 percent of primary energy consumption. These figures are not expected to change in the near future, due to suspended reactor construction in many countries. Nevertheless, world energy needs continue to grow: the planet's population already exceeds five billion and is forecast to reach ten billion by the middle of the next century. Most less developed countries have a very low rate of energy consumption and, even though some savings can be made in industrialized countries, it will become increasingly difficult to satisfy needs using fossil fuels only. Furthermore, there has been no recent breakthrough in the energy landscape. The physical feasibility of the other great hope of nuclear energy, fusion, has yet to be proved; once this has been done, it will be necessary to solve technological problems and to assess economic viability. Although it is more ever necessary to pursue fusion programs, there is little likelihood of industrial applications being achieved in the coming decades. Coal and fission are the only ways to produce massive amounts of energy for the next century. Coal must overcome the pollution problems inherent in its use; fission nuclear power has to gain better public acceptance, which is obviously colored by safety and waste concerns. Most existing reactors were commissioned in the 1970s; reactor lifetime is a parameter that has not been clearly established. It will certainly be possible to refurbish some to extend their operation beyond the initial target of 30 or 40 years. But normal advances in technology and safety requirements will make the operation of the oldest reactors increasingly difficult. It becomes necessary to develop new generations of nuclear reactors, both to replace older ones and to revive plant construction in their countries that are not yet equipped or that have halted their

  2. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  3. Operating US power reactors

    International Nuclear Information System (INIS)

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of Dec. 31, 1986, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (October, November, and December 1986) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. They are defined as follows: In addition to the tabular data, this article discusses significant occurrences and developments that affected licensed US power reactors during this reporting period. It includes, but is not limited to, changes in operating status, regulatory actions and decisions, and legal actions involving the status of power reactors. We do not have space here for routine problems of operation and maintenance, but such information is available at the Nuclear Regulatory Commission (NRC) Public Document Room, 1717 H Street, NW, Washington, DC 20555. Some significant operating events are summarized elsewhere in this section in the article ''Selected Safety-Related Events,'' and a report on activities relating to facilities still in the construction process is given in the article ''Status of Power-Reactor Projects Undergoing Licensing Review'' in the last section of each issue of this journal. The reader's attention is also called to the regular feature ''General Administrative Activities,'' which deals with more general aspects of regulatory and legal matters that are not covered elsewhere in the journal

  4. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  5. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  6. Experimental reactor physics

    International Nuclear Information System (INIS)

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

  7. Diagnostics for hybrid reactors

    International Nuclear Information System (INIS)

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  8. Perspectives on reactor safety

    International Nuclear Information System (INIS)

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  9. Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Mr. Baron says the administration's effort to terminate the Clinch River Breeder Reactor (CRBR) project is symptomatic; they have also placed restrictions on fusion, coal, solar, and other areas of energy development in which technological advances are held back in order to force conservation. Because the breeder reactor, unlike solar and fusion energy, is both economically and technically feasible, a demonstration plant is needed. The contentions that the CRBR design is obsolete, that its proposed size is inappropriate, or that plutonium can be diverted for weapons proliferation are argued to be invalid. Failure to complete the CRBR will have both economic and national security repercussions

  10. Netherlands Interuniversity Reactor Institut

    International Nuclear Information System (INIS)

    This is the annual report of the Interuniversity Reactor Institute in the Netherlands for the Academic Year 1977-78. Activities of the general committee, the daily committee and the scientific advice board are presented. Detailed reports of the scientific studies performed are given under five subjects - radiation physics, reactor physics, radiation chemistry, radiochemistry and radiation hygiene and dosimetry. Summarised reports of the various industrial groups are also presented. Training and education, publications and reports, courses, visits and cooperation with other institutes in the area of scientific research are mentioned. (C.F.)

  11. Reactor Materials Research

    International Nuclear Information System (INIS)

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  12. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  13. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

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

  14. Decay of reactor neutrinos

    OpenAIRE

    Vogel, P.

    1984-01-01

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Gösgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit tc.m. / mν is 1-3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1...

  15. Small mirror fusion reactors

    International Nuclear Information System (INIS)

    Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

  16. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  17. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  18. Reactor PIK construction

    International Nuclear Information System (INIS)

    The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

  19. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  20. Reactor gamma spectrometry: status

    International Nuclear Information System (INIS)

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

  1. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

  2. Space-time reactor kinetics for heterogeneous reactor structure

    International Nuclear Information System (INIS)

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods

  3. Experimental studies on sodium fires. Application to reactors

    International Nuclear Information System (INIS)

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

  4. Safety of civil basic nuclear installations other than reactors of the French nuclear fleet. Lessons learned by the IRSN and significant events declared in 2013 and 2014

    International Nuclear Information System (INIS)

    Published every two years, this report concerns installations related to fuel nuclear cycle, research laboratories and reactors, installations for waste processing, warehousing or storage of nuclear wastes, as well as installations which have been definitely stopped or are being remediated or dismantled. After a global and synthetic overview of these issues (highlights, lessons learned from significant events, evolution of the number of events, lessons learned from safety-re-examinations), the report first presents the legal and regulatory context, the 82 concerned installations, and aspects related to their safety. The second part proposes an analysis of significant events which occurred in 2013 and 2014: aspects related to event declaration, assessment of events and main noticed trends with respect to previous years, transverse analysis of the main types of events and evolutions with respect to previous years. The third part reports some important events which occurred in the Fontenay-aux-Roses CEA centre, in the Atalante installation, on the Tricastin site, and in the UP3-A plant of Areva NC establishment in La Hague. Some transverse topics are then addressed, i.e. safety re-examinations of various installations (the UP3-A plant in La Hague, the MELOX plant, the CEA Cadarache solid waste processing station, and the Phenix power plant)

  5. Risk prevention during reactor shutdown

    International Nuclear Information System (INIS)

    During reactor shutdown potential risks are issued of a number of maintenance operations. In this text we analyse these operations and give the modifications of technical specifications to ameliorate the reactor safety. 4 figs

  6. New fast-reactor approach

    International Nuclear Information System (INIS)

    The design parameters for a 1000 MW LMFBR type reactor are presented. The design requires the multiple primary coolant pumps and heat exchangers to be located around the core within the reactor vessel

  7. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  8. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  9. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  10. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  11. The IR-8 reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ryazantsev, E.P.; Egorenkov, P.M.; Yashin, A.F. [Reactor Technology and Materials Research Inst. of RRC ' KI' , Moscow (Russian Federation)

    1997-07-01

    At the Russian Research Center 'Kurchatov Institute' (RRC 'KI') the IR-8 reactor commissioning was carried out in 1981. The reactor was developed in return for earlier existing at RRC 'KI' of the IRT-M reactor (modernized IRT reactor, constructed in 1957). The IRT-M reactor was used for investigations in nuclear physics, solid state physics, radiation chemistry, biology as well as to produce isotopes. Under developing the IR-8 reactor the IRT biological shielding with beam tubes and its process systems were used. The IR-8 reactor creation was founded on application developed by then new fuel assemblies (FA) of IRT-3M type, having two times as great surface of heat transfer and 1.75 times higher U-235 load than the FA of the IRT-2M type, which were used in IRT-M reactor. (author)

  12. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  13. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  14. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  15. VVER and RBMK reactors

    International Nuclear Information System (INIS)

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

  16. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  17. Studies on reactor physics

    International Nuclear Information System (INIS)

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  18. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  19. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  20. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  1. Nuclear reactor container

    International Nuclear Information System (INIS)

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  2. ICF tritium production reactor

    International Nuclear Information System (INIS)

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g

  3. Fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1989-01-01

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

  4. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  5. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  6. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarizes the fast reactor research carried out by ECN during the period covering the year 1980. This research is mainly concerned with the cores of sodium-cooled breeders, in particular the SNR-300, and its related safety aspects. It comprises six items: A programme to determine relevant nuclear data of fission- and corrosion-products; A fuel performance programme comprising in-pile cladding failure experiments and a study of the consequences of loss-of-cooling and overpower; Basic research on fuel; Investigation of the changes in the mechanical properties of austenitic stainless steel DIN 1.4948 due to fast neutron doses, this material has been used in the manufacture of the reactor vessel and its internal components; Study of aerosols which could be formed at the time of a fast reactor accident and their progressive behaviour on leaking through cracks in the concrete containment; Studies on heat transfer in a sodium-cooled fast reactor core. As fast breeders operate at high power densities, an accurate knowledge of the heat transfer phenomena under single-phase and two-phase conditions is sought. (Auth.)

  7. Fusion reactor materials

    International Nuclear Information System (INIS)

    At the Belgian Nuclear Research Centre SCK-CEN, activities related to fusion focus on environmental tolerance of opto-electronic components. The objective of this program is to contribute to the knowledge on the behaviour, during and after neutron irradiation, of fusion-reactor materials and components. The main scientific activities for 1997 are summarized

  8. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  9. Cermet fuel reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

    1987-09-01

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

  10. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs

  11. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  12. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

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

  13. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  14. Fusion reactor materials

    International Nuclear Information System (INIS)

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

  15. Department of reactor technology

    International Nuclear Information System (INIS)

    The activities of the Department of Reactor Technology at Risoe during 1980 are described. The work is presented in three chapters: General Information on the Department, Summary of the Department's Development during 1980, and Activities of the Department. Lists of staff, publications, computer programs, and test facilities are included. (author)

  16. The AP1000 reactor

    International Nuclear Information System (INIS)

    The design of the AP1000 reactor began 20 years ago when Westinghouse launched the AP600 reactor project. In fact by re-assessing AP600's safety margins Westinghouse realized that the its power output could be raised without putting at risk its safety standard. The AP1000 was born, it yields 1100 MWe. The main AP1000's design features is its passive safety (particularly after the Fukushima accident) and its modularity. The passive safety of the AP1000 implies: -) no humane intervention needed for 72 hours at least after the incident; -) no necessity for redundant complex safety systems. The modularity means that the plant, the reactor and other buildings are constructed from a choice of 300 modular units. These units can be built off-site and fit together on site. The modularity allows more construction activities to be led simultaneously and more chances to cope with the construction schedule. The NRC has approved the operation license for 30 years of the first AP1000 being built in the Usa (Vogtle plant in Georgia). 4 AP1000 are being built in China (Sanmen and Haiyang sites) and 6 others are planned in the Usa. Westinghouse is convinced that the AP1000's passive safety makes it more attractive. Let us not forget that Westinghouse was at the origin of the concept of pressurized water reactors, an idea adopted for half the nuclear power stations in the world and for all the plants now active in France. (A.C.)

  17. Fast reactor programme

    International Nuclear Information System (INIS)

    This progress report summarises the fast reactor research carried out at the Netherlands Energy Research Centre during the year 1981. The neutron and fission product cross sections of various isotopes have been evaluated. In the fuel performance programme, some preliminary results are given and irradiation facilities described. Creep experiments on various stainless steel components are reported

  18. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  19. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    The general design, characteristics and parameters of TRIGA reactors and fuel are described. This is a training module with the learning objectives: to understand the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and realize the differences between TRIGA fuels and other more traditional. 10 figs., 6 tabs. (nevyjel)

  20. SNAP Nuclear Space Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  1. Fast breeder reactor

    International Nuclear Information System (INIS)

    This paper outlined the present status of FBR development in six countries and reviewed Japanese activities on FBR development. Joyo experimental FBR has accumulated a lot of technical data including irradiation tests of advanced fuels and was now long shut down due to the partial obstruction of rotating plug movement. Monju prototype FBR reactor experienced a sodium leakage in its secondary heat transfer system during performance tests in December 1995 and had been shut down until May 2010. Feasibility study on commercialized FBR cycle system ended in March 2006 and proposed the concept of commercialized FBR cycle technologies. In order to plan a demonstration reactor, research and development of innovative technologies are conducted as the FaCT (Fast Reactor Cycle Technology Development) Project. In connection with the results of this research and development, a 5-party council of Japan was established to discuss processes of demonstration and commercialization of FBR cycle systems in Japan. Joint efforts were made for a demonstration reactor to be committed in 2015, in addition to start operation around 2025 aiming at the commercialization of FBR before 2050. (T. Tanaka)

  2. The Chernobyl reactor accident

    International Nuclear Information System (INIS)

    The documentation abstracted contains a complete survey of the broadcasts transmitted by the Russian wire service of the Deutsche Welle radio station between April 28 and Mai 15, 1986 on the occasion of the Chernobyl reactor accident. Access is given to extracts of the remarkable eastern and western echoes on the broadcasts of the Deutsche Welle. (HP)

  3. Cermet fuel reactors

    International Nuclear Information System (INIS)

    Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. The concept evolved in the 1960's with the objective of developing a reactor design which could be used for a wide range of mobile power generation systems including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests and in-reactor irradiation tests using cermet fuel were carried out by General Electric in the 1960's as part of the 710 Development Program and by Argonne National laboratory in a subsequent activity. Cermet fuel development programs are currently underway at Argonne National laboratory and Pacific Northwest Laboratory as part of the Multi-Megawatt Space Power Program. Key features of the cermet fueled reactor design are 1) the ability to achieve very high coolant exit temperatures, and 2) thermal shock resistance during rapid power changes, and 3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, there is a potential for achieving a long operating life because of 1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and 2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core

  4. Reactor power measuring device

    International Nuclear Information System (INIS)

    The device of the present invention comprises a γ-thermometer disposed in a BWR type reactor, a first amplifier for amplifying the output thereof, a fission ionization chamber disposed in the reactor separately from the γ-thermometer, a second amplifier for amplifying the output thereof, a differential circuit for differentiating the output signal of the second amplifier and a first adding circuit for adding an output signal of the differential circuit and an output signal of the first amplifier. Alternatively, a γ-ray self-powered neutron detector may be disposed instead of the fission ionization chamber. A second adding circuit is also disposed for adding the output signals of plurality of differentiation circuits and inputting the result to the first adding circuit. A sensitivity controller is disposed upstream of the first adder for controlling the sensitivity of the fission ionization chamber. Then, even if time delay should be caused in the γ-thermometer, output signals with good time response characteristic can be obtained by using signals of LPRM or SPND, and currently changing output of the reactor can be measured accurately to provide an effect on the improvement of the safety and operation controllability of the reactor. (N.H.)

  5. Surveying and Comparing the Arco Dei Gavi and its Historical Wooden Maquette

    OpenAIRE

    Francesco Guerra; Paolo Vernier

    2011-01-01

    Actually geometrics’ science offers new opportunities and interesting applications in the field of Cultural Heritage. These applications are strictly related to preservation, restoration but even to cataloging and reproducing a monument that no longer has its original integrity.  The possibility of obtaining 3D data, of such a model close to reality, enables us to realize studies that sometimes are too complex or impossible. The paper will describe the study of a monumental arch, the Arco dei...

  6. Surveying and Comparing the Arco Dei Gavi and its Historical Wooden Maquette

    Directory of Open Access Journals (Sweden)

    Francesco Guerra

    2011-12-01

    Full Text Available Actually geometrics’ science offers new opportunities and interesting applications in the field of Cultural Heritage. These applications are strictly related to preservation, restoration but even to cataloging and reproducing a monument that no longer has its original integrity.  The possibility of obtaining 3D data, of such a model close to reality, enables us to realize studies that sometimes are too complex or impossible. The paper will describe the study of a monumental arch, the Arco dei Gavi, built in Verona during the I sec. A.C., that was destroyed in 1805 by the Napoleonic army, and its wooden model that was realized in 1813 and it has a very important role concerning the monument’s reconstruction. The purpose is to realize two threedimensional models which can be comparable to each other, two models with recognizable differences, similarities and discontinuities about shapes and single elements that compose the monument. It should also be noted that some original parts of the monument have not been relocated but are badly preserved in a museum: the 3D digital model helps to identify these parts in their original location. The main steps of the work can be summarized in: collecting the historical documentation of Arco dei Gavi and its representations; identifying proper instruments (laser scanning and photogrammetric hardware and software; surveying the Arch and its wooden model; identifying a unique and shared reference system; comparing both digital models related to the same scale; choosing a three-dimensional representation to emphasize the results; reallocation of outstanding pieces (virtual anastylosis.

  7. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  8. Reactor physics problems on HCPWR

    International Nuclear Information System (INIS)

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  9. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  10. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

  11. Reactor regulating and protection system for a light water reactor

    International Nuclear Information System (INIS)

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  12. The reactor Pegase

    International Nuclear Information System (INIS)

    The reactor Pegase is designed for testing fuel elements for gas-cooled power reactors. Experience has shown that the classical multi-purpose test reactors are not well adapted to these tests. On another side, the introduction of these test elements into the existing power reactors involves numerous problems, which limits their interest. Pegase, which is designed to satisfy these experimental needs, is composed of a parallelepipedal core of enriched Uranium, moderated and cooled by pressurized water. This core is used as a neutron source for eight autonomous loops, containing the elements to be tested, and situated around the core. The core and the eight loops are immersed in a irradiation pool. The loops are placed on the bottom of the pool so, it is possible to move a loop away from the core, or to remove it from the pool without interfering with the operation of the other loops. The irradiation conditions are adjusted, making the synthesis of the following development works. - Experimental studies on Peggy, a zero power critical facility, mock up of Pegase in operation since 1961: measurements of neutron flux level, radial and axial fly distributions on the experiments. Effect of burnable poisons and of movements of the control rods; adjustment of devices (reflectors, screens etc..) needed for optimum performances. - Experimental work on two prototype autonomous loops, heated electrically to the nominal operating power (in operation since 1961): development of the thermodynamic measurements, thermal balances parameters for control of the operating conditions, natural convection. - Studies on Pegase operating under power; thermodynamic measurements on the core circuits on the independent loop circuits; neutronic measurements, etc... The reactor Pegase went critical on the 4. of April 1963 and reached the nominal power of 30 MW on the 28. of May 1963. (authors)

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

    International Nuclear Information System (INIS)

    The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in 235U) fuels (U3Si2 for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has been performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAlx/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm3. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% 235U) by the Minerve-dedicated interpretation tool: PIMS. The effect of

  14. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  15. Reactor physics activities in Japan

    International Nuclear Information System (INIS)

    This report reviews the research activity in reactor physics field in Japan during July, 1992 - July, 1993. The review was performed in the following fields : nuclear data evaluation, calculational method development, fast reactor physics, thermal reactor physics, advanced core design, fusion reactor neutronics, nuclear criticality safety, shielding, incineration of radioactive nuclear wastes, noise analysis and control and national programs. The main references were taken from journals and reports published during this period. The research committee of reactor physics is responsible for the review work. (author)

  16. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

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

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  18. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...... against Methanothrix soehngenii or Methanothrix CALS-I in any of the thermophilic biogas reactors examined. Studies using 2-14C-labeled acetate showed that at high concentrations (more than approx. 1 mM) acetate was metabolized via the aceticlastic pathway, transforming the methyl-group of acetate...

  19. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIMtm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  20. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  1. Feasible reactor power cutback logic development for an integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Soon-Kyoo [KHNP Co., Ltd., Uljin-gun, Gyeong-buk (Korea, Republic of); Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok [Korea Atomic Energy Research Institute (KAERI), Daedeokdaero, Yuseong, Daejeon (Korea, Republic of)

    2013-07-15

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  2. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  3. Elk River Reactor dismantling

    International Nuclear Information System (INIS)

    The dismantling program was carried out in three overlapping phases: the planning phase which included the preliminary planning and selection of the dismantling approach, the dismantling phase which included all work performed to remove the reactor facility and restore the site to its pre-reactor condition, and the closeout phase which included the final site survey and efforts necessary to terminate the AEC license and contract. Of particular interest was the use of a remotely operated plasma cutting torch to section the pressure vessel internals, the pressure vessel and the outer thermal shield, the use of explosives in removal of the biological shield and the method of establishment of the criteria for material disposal

  4. Selecting reactor operator trainees

    International Nuclear Information System (INIS)

    Reactor operator trainee selection tends to be more effective if tailored to a utility's unique needs, and offers the organization a better chance for compliance with Federal regulations than if selection methods are adopted without benefit of local research. The costs of operator training range from $50,000 to $100,000. The test validity relative to a variety of training grades and performance measures is reviewed. Of interest is the degree to which tests differentiate reactor operators with respect to simulator training grades and performance in simulator operation; forms of evaluation which have become fairly standard throughout the power industry. The tests administered to each individual were selected because of their presumed relevance to training grades, and the aptitude measures are intended to assess an individual's potential to benefit from training. Tests, availability, form, the abilities they measure, and the time limit are described. (MCW)

  5. Embattled breeder reactor

    International Nuclear Information System (INIS)

    A commercial fuel-cloning machine, a nuclear breeder reactor, is yet to produce electricity in the United States. It is expensive in capital and fuel costs, its fuel that must be reprocessed can become a link to nuclear weapons manufacture, and its safety is no greater than conventional nuclear reactors. The breeder has had on-again/off-again administrative support from Washington. Opponents worry about escalating costs and failure to develop alternatives like solar energy. Proponents say fossil-fuel depletion will eventually force long-term renewable resources such as the breeder anyway. Some who share parts of both views oppose present policy regarding the Clinch River Breeder demonstration plant specifically. The correct choices on breeder concept development and commercialization will be known in 2050. 3 figures

  6. Robot for reactor dismantling

    International Nuclear Information System (INIS)

    Purpose: To enable to attain the operation on a cylindrical coordinate system thereby performing dismantling operation exactly and at a high reliability. Constitution: A reactor dismantling robot is suspended by ropes by an elevating device to the inside of reactor shielding walls. The robot has a fixed portion having a plurality of legs abutting against the inner surface of the shieling walls while extending and shrinking radially in the horizontal direction and an arm portion having an operation arm disposed with a shielding wall breaking operation device. The arm portion is disposed with a mechanism for vertically moving the operation arm and a mechanism for forwarding and backwarding the operation arm in the horizontal direction and the arm portion itself is constituted so as to be rotatable around a vertical axis. (Seki, T.)

  7. Compact fusion reactors

    CERN Document Server

    CERN. Geneva

    2015-01-01

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

  8. The Pegase reactor loops

    International Nuclear Information System (INIS)

    After 4 years operation, experimentation and maintenance of the gas loops built especially for the nuclear fuel testing reactor Pegase, it appears desirable not only to gather together in a single document the essential characteristics and particularities of these devices and of their associated equipment, but also to give the reasons for the technical modifications and the way in which they were carried out; this is done here by the persons themselves who were responsible, day after day, for operating these loops. This essentially practically experience thus complements the careful research and preliminary testing carried out on these loops or on their prototypes. It should be of interest to those who deal with problems concerned with the design or operation of irradiation loops in experimental reactors or of similar equipment. (authors)

  9. Spherical torus fusion reactor

    Science.gov (United States)

    Peng, Yueng-Kay M.

    1989-01-01

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  10. Decay of reactor neutrinos

    International Nuclear Information System (INIS)

    We consider the decay of massive neutrinos which couple to electrons and are, therefore, produced in nuclear reactors. Lifetime limits for the γ and electron-positron decay modes of these neutrinos are deduced from the experimental limit on the singles count rate in the detector used to study neutrino oscillations at the Goesgen reactor. The dominantly coupled neutrinos are light, and their invariant-lifetime limit t/sup c.m.//m/sub ν/ is 1--3 sec/eV. The subdominantly coupled heavy neutrinos with mass 1--4 MeV could decay into electron-positron pairs. These pairs were not observed, and from the absence of such a signal we deduce restrictions on the corresponding mixing parameters

  11. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  12. Fast reactor database

    International Nuclear Information System (INIS)

    This publication contains detailed data on liquid metal cooled fast reactors (LMFRs), specifically plant parameters and design details. Each LMFR power plant is characterized by about 400 parameters, by design data and by relevant materials. The report provides general and detailed design characteristics including structural materials, data on experimental, demonstration, prototype and commercial size LMFRs. The focus is on practical issues that are useful to engineers, scientists, managers and university students and professors. The report includes updated information contained in IAEA previous publications on LMFR plant parameters: IWGRF/51 (1985) and IWGFR/80 (1991) and reflects experience gained from two consultants meetings held in Vienna (1993,1994). This compilation of data was produced by members of the IAEA International Working Group on Fast Reactors (IWGFR)

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

    International Nuclear Information System (INIS)

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

  14. Gaseous fuel reactor research

    Science.gov (United States)

    Thom, K.; Schneider, R. T.

    1977-01-01

    The paper reviews studies dealing with the concept of a gaseous fuel reactor and describes the structure and plans of the current NASA research program of experiments on uranium hexafluoride systems and uranium plasma systems. Results of research into the basic properties of uranium plasmas and fissioning gases are reported. The nuclear pumped laser is described, and the main results of experiments with these devices are summarized.

  15. Neutrino physics at reactors

    International Nuclear Information System (INIS)

    Reviewing experiments with neutrinos from reactors seems at first to be a simple task, since there were only few. But almost all of them address fundamental questions in particle physics and are of great relevance. This paper reports on these experiments which made use of some of the most sophisticated techniques available at the time they were designed. In these two respects new proposed experiments re in the tradition of the older ones

  16. Decommissioning of research reactors

    International Nuclear Information System (INIS)

    Research reactors of WWR-S type were built in countries under Soviet influence in '60, last century and consequently reached their service life. Decommissioning implies removal of all radioactive components, processing, conditioning and final disposal in full safety of all sources on site of radiological pollution. The WWR-S reactor at Bucuresti-Magurele was put into function in 1957 and operated until 1997 when it was stopped and put into conservation in view of decommissioning. Presented are three decommissioning variants: 1. Reactor shut-down for a long period (30-50 years) what would entail a substantial decrease of contamination with lower costs in dismantling, mechanical, chemical and physical processing followed by final disposal of the radioactive wastes. The drawback of this solution is the life prolongation of a non-productive nuclear unit requiring funds for personnel, control, maintenance, etc; 2. Decommissioning in a single stage what implies large funds for a immediate investment; 3. Extending the operation on a series of stages rather phased in time to allow a more convenient flow of funds and also to gather technical solutions, better than the present ones. This latter option seems to be optimal for the case of the WWR-S Research at Bucharest-Magurele Reactor. Equipment and technologies should be developed in order to ensure the technical background of the first operations of decommissioning: equipment for scarification, dismantling, dismemberment in a highly radioactive environment; cutting-to-pieces and disassembling technologies; decontamination modern technologies. Concomitantly, nuclear safety and quality assurance regulations and programmes, specific to decommissioning projects should be implemented, as well as a modern, coherent and reliable system of data acquisition, recording and storing. Also the impact of decommissioning must be thoroughly evaluated. The national team of specialists will be assisted by IAEA experts to ensure the

  17. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    An improvement in the construction of liquid metal cooled nuclear reactors of the kind in which the fuel assembly is submerged in a pool of coolant contained by a primary vessel housed in a concrete vault, is described. In this modification the roof of the vault carries heat exchangers immersed in the pool of coolant, the lower ends of which are hydraulically damped against oscillation caused by seismic disturbances. (U.K.)

  18. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  19. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

  20. Halden reactor project

    International Nuclear Information System (INIS)

    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