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Sample records for epr type reactor

  1. The EPR reactor

    International Nuclear Information System (INIS)

    Lacoste, A.C.; Dupuy, Ph.; Gupta, O.; Perez, J.R.; Emond, D.; Cererino, G.; Rousseau, J.M.; Jeffroy, F.; Evrard, J.M.; Seiler, J.M.; Azarian, G.; Chaumont, B.; Dubail, A.; Fischer, M.; Tiippana, P.; Hyvarinen, J.; Zaleski, C.P.; Meritet, S.; Iglesias, F.; Vincent, C.; Massart, S.; Graillat, G.; Esteve, B.; Mansillon, Y.; Gatinol, C.; Carre, F.

    2005-01-01

    This document reviews economical and environmental aspects of the EPR project. The following topics are discussed: role and point of view of the French Nuclear Safety Authority on EPR, control of design and manufacturing of EPR by the French Nuclear Safety Authority, assessment by IRSN of EPR safety, research and development in support of EPR, STUK safety review of EPR design, standpoint on EPR, the place of EPR in the French energy policy, the place of EPR in EDF strategy, EPR spearhead of nuclear rebirth, the public debate, the local stakes concerning the building of EPR in France at Flamanville (Manche) and the research on fourth generation reactors. (A.L.B.)

  2. EPR (European Pressurized Reactor)

    International Nuclear Information System (INIS)

    2015-01-01

    This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

  3. The EPR in a few words: all you need to know about the EPR nuclear reactor

    International Nuclear Information System (INIS)

    2009-01-01

    After a brief presentation of the EPR (European - or Evolutionary - Pressurized Reactor) type nuclear reactor, this paper, proposed by the collective group 'Stop EPR', develops the following points: EPR is as dangerous as other reactors; EPR flouts democracy; France's energy demand do not need the construction of EPRs; the construction of EPRs is not a factor of economical and social development; EPR should not be constructed neither in France nor elsewhere and the present building sites should be cancelled; the EPR will not help France to increase its energy independence and protect itself from oil price increases; choosing the EPR is incompatible with the large investments to be made in energy conservation and renewable energies; the EPR is not a solution to climate change; the VHV line corridor that will starts at Flamanville is not justified and poses risks to the environment and public health

  4. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

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

  5. The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

  6. Towards EPR (European pressurized reactor)

    International Nuclear Information System (INIS)

    Anon.

    2003-01-01

    According to the French industry minister, it is nonsense continuing delaying the construction of an EPR prototype because France needs it in order to renew timely its park of nuclear reactors. The renewing is expected to begin in 2020 and will be assured with third generation reactors like EPR. A quick launching of the EPR prototype is necessary to have it being in service by 2012, the feedback operating experience that will be accumulated over the 8 years that will follow will be necessary to optimize the industrial version and to have it ready by 2020. The EPR reactor has indisputable assets: modern, safer, more competitive and it will produce less wastes than present nuclear reactors. The construction cost of an EPR prototype is estimated to 3 milliard Euros and the nuclear industry operators propose to finance it completely. The EPR prototype does not jeopardize the ambitious French program about renewable energy sources, France is committed to produce 21% of its electricity from renewable energies by 2010 and 10 milliard Euros will be invested over this period on wind energy. Nuclear energy and alternative energies must be considered as 2 aspects of a diversified energy policy. (A.C.)

  7. Not a mystery. Inner containment of the pressurized water reactor (EPR trademark type)

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Wienand, Burkhard; Krumb, Christian [AREVA NP GmbH (Germany)

    2012-11-01

    The containment of the advanced pressurized water reactor EPR trademark type is developed on the basis of the French nuclear power plant operational experience and consists of - The reinforced outer containment structure, designed to withstand external hazards (e.g. APC), - The pre-stressed inner containment structure, designed to bear the loads resulting from internal hazards (LOCA), - The steel liner, designed to provide leak tightness resulting from internal hazards. The main advantage of the pre-stressed inner containment design is that the structure remains in linear-elastic behavior during the whole life-time. Even in case of postulated design accidents (LOCA) concrete tensile strains are strongly limited. Due to pre-stressing the concrete structure remains practically free of cracks. Due to pre-stressing the leak tightness ensuring steel liner, embedded into the inner concrete shell, is exposed to more favorable compression loads. In addition to detailed calculations several test programs have been performed to verify and confirm the predicted behavior in normal operation and in accident condition. (orig.)

  8. EPR by Areva. EPR the 1600+ MWe reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system.

  9. EPR by Areva. EPR the 1600+ MWe reactor

    International Nuclear Information System (INIS)

    2008-01-01

    This brochure presents the GEN III+ EPR reactor designed by the Areva and Siemens consortium. The EPR reactor is a direct descendent of the well-proven N4 and KONVOI reactors, the most modern reactors in France and Germany. The EPR was designed by teams from KWU/Siemens and Framatome, EDF in France and the major German utilities, working in collaboration with both French and German safety authorities. The EPR integrates the results of decades of R and D programs, in particular those performed by the CEA (French Atomic Energy Commission) and the Karlsruhe Research Center in Germany. The EPR benefits from the experience of several thousand reactor-years of operation of pressurized water reactor technology. This experience has put 87 AREVA PWRs online throughout the world. Innovative Features: - An outer shell covering the reactor building, the spent fuel building and two of the four safeguard buildings provides protection against large commercial or military aircraft crash. - A heavy neutron reflector that surrounds the reactor core lowers uranium consumption. - An axial economizer inside the steam generator allows a high level of steam pressure and therefore high plant efficiency. - A core catcher allows passive collection and retention of the molten core should the reactor vessel fail in the highly unlikely event of a core melt. - A digital technology and a fully computerized control room with an operator friendly man-machine interface improve the reactor protection system

  10. Radiation protection issues for EPR reactor

    International Nuclear Information System (INIS)

    Miniere, D.; Le Guen, B.; Beneteau, Y.; Le Guen, B.

    2008-01-01

    As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection Radiation Protection aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing Radiation Protection-significant activities has been to include all involved stakeholders - designers, licensee and contractor companies - in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The Radiation Protection target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 Man Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the E.P.R.'s reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), Radiation Protection studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the

  11. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  12. EPR: The reactor generation for the 21st century. SFEN-KTG congress 'The EPR Project'', Strasbourg, 13-14 Nov. 1994

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The foundation of the subsidiary in 1989, Nuclear Power International (NPI), brought together as partners the two formerly competing companies Siemens/KWU and Framatome who now are planning the novel reactor type within the framework of the joint EPR project. The project goal is to develop a novel French-German PWR reactor type for power generation, named European Pressurized Water Reactor (EPR). The article summarizes the main results of papers and discussions of the congress. (orig./UA) [de

  13. EPR by AREVA. An evolutionary reactor

    International Nuclear Information System (INIS)

    Horstmann, Marion

    2010-01-01

    The EPR development goals are as follows: 1. Evolutionary design to fully capitalize on the design, construction and operating experience based on the 86 AREVA's PWR operating worldwide; 2. Enhanced Safety compared to operating PWRs: reduce core damage frequency (CDF), accommodate severe accidents with no long-term population effect, Withstand large airplane crash (APC); 3. High availability; 4. Simplified operation and maintenance; and 5. Generation cost at least 10 % lower than 1500 MWe series in operation.The design builds on the achievements of the N4 and Konvoi reactors. The main plant data are tabulated. The PWR structure is shown as an example of the stepwise improvement. Focus of the presentation is on the construction techniques, supply chain, and project delivery. (P.A.)

  14. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  15. Meeting the next generation PWR safety requirements: The EPR Reactor

    International Nuclear Information System (INIS)

    Salhi, Othman

    2008-01-01

    The development process pursued the harmonization of technical solutions and the integration of all the lessons learned from earlier nuclear plants built by both vendors. As far as safety more specifically is concerned, the basic choice for the EPR was to adopt an evolutionary approach based on experience feedback from the reactors built by Areva, which at the time already amounted to nearly 100. This philosophy makes today's Areva EPR the natural descendant of the most advanced French N4 and German Konvoi power reactors currently in operation. EPR design choices affecting safety were motivated by a continuous quest for higher levels of safety. A two-fold approach was followed: 1. improvement of the measures aimed at further reducing the already very low probability of core melt 2. incorporation of measures aimed at further limiting the consequences of a severe accident, in the knowledge that its probability of occurrence has been considerably reduced. Through its filiations with French N4 and German Konvoi power reactors, the EPR benefits from the uninterrupted, evolutionary innovation process that has supported the development of PWRs since their introduction into the market place. This is especially true for safety where the EPR brings a unique combination of both tried and tested and innovative features that further improve the prevention of severe accidents and their mitigation

  16. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

  17. EPR, a GEN 3 Reactor providing a competitive electricity cost

    International Nuclear Information System (INIS)

    Salhi, Othman

    2006-01-01

    Since the very beginning of the development of what was to become the EPR, several European entities were involved. The French and German safety authorities expressed that reinforced safety was compulsory. Additional measures were then included to prevent the occurrence of events likely to damage the core, and reduce the possibility of exposure of operating and maintenance personnel. However, not with standing these safety related features resulting from the requirements of the safety authorities, we will focus today on another group of entities that were key players in EPR development: the Utilities. The Utilities voiced their need for a competitive electricity produced and a competitive nuclear reactor. The tradeoff was then to reach both targets in a unique product: a safer and more competitive NPP. Today, the EPR presents features that enable our clients to compete with the cheapest fossil-based electricity production plants. Increased thermal efficiency is obtained both through a higher steam pressure and through careful optimization of the secondary system thermal cycle

  18. 75 FR 12804 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR; Notice...

    Science.gov (United States)

    2010-03-17

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR; Notice of Meeting The ACRS U.S. Evolutionary Power Reactor (EPR) Subcommittee will hold a... Chapters 11 and 16 of the Safety Evaluation Report with Open Items associated with the U.S. EPR Design...

  19. 75 FR 7633 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR; Notice...

    Science.gov (United States)

    2010-02-22

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR; Notice of Meeting The ACRS U.S. Evolutionary Power Reactor (EPR) Subcommittee will hold a....S. EPR Design Certification Document Review. The Subcommittee will gather information, analyze...

  20. 75 FR 27841 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR

    Science.gov (United States)

    2010-05-18

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on EPR The ACRS Subcommittee on EPR will hold a meeting on May 21, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting will be open to public attendance. The agenda for...

  1. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  2. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  3. China: the construction of the Taishan number 1 EPR reactor reaches a major milestone with the setting of the vessel

    International Nuclear Information System (INIS)

    Anon.

    2012-01-01

    This article describes some characteristics of the reactor vessel which has recently been introduced in the building of the number 1 EPR reactor of Taishan in China; the construction of this EPR reactor is coordinated by EDF, CGNPC and Areva. It also briefly describes how this introduction has been performed, and evokes the fabrication quality issues and tests

  4. Characteristics and economy of the European reactor of pressurized water (EPR); Caracteristicas y economia del reactor europeo de agua a presion (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jov@nuclear.inin.mx

    2005-07-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  5. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  6. New methods for data processing and designing the EPR European reactor

    International Nuclear Information System (INIS)

    Bogoslovskij, A.V.

    1999-01-01

    The basic parameters of the reactor in the assumption of its capacity, equalling 1500 MW and meeting the requirements of the European consumers, are presented. The experience in operating the nuclear facilities in France and Germany during above 1000 reactor/years is used by designed the EPR. The EPR safety systems are based on the maximum simplicity by application of the stand by principle and provision for the redundancy with separated functions. Application of standardization both for the NPP as a whole and individual equipment makes it possible to decrease the construction costs, as well as the expenditures on electrical energy production and the NPP personnel training [ru

  7. Minimization of actinide waste by multi-recycling of thoriated fuels in the EPR reactor

    Directory of Open Access Journals (Sweden)

    Nuttin A.

    2012-02-01

    Full Text Available The multi-recycling of innovative uranium/thorium oxide fuels for use in the European Pressurized water Reactor (EPR has been investigated. If increasing quantities of 238U, the fertile isotope in standard UO2 fuel, are replaced by 232Th, then a greater yield of new fissile material (233U is produced during the cycle than would otherwise be the case. This leads to economies of natural uranium of around 45% if the uranium in the spent fuel is multi-recycled. In addition we show that minor actinide and plutonium waste inventories are reduced and hence waste radio-toxicities and decay heats are up to a factor of 20 lower after 103 years. Two innovative fuel types named S90 and S20, ThO2 mixed with 90% and 20% enriched UO2 respectively, are compared as an alternative to standard uranium oxide (UOX and uranium/plutonium mixed oxide (MOX fuels at the longest EPR fuel discharge burn-ups of 65 GWd/t. Fissile and waste inventories are examined, waste radio-toxicities and decay heats are extracted and safety feedback coefficients are calculated.

  8. Optimisation of radiation protection for the new european pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Miniere, D.; Beneteau, Y.; Le Guen, Bernard

    2008-01-01

    Full text: As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection (RP) aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing RP-significant activities has been to include all involved stake holders -designers, licensee and contractor companies- in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The RP target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 man.Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the EPR 's reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), RP studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the accessible area can be posted as a green area (dose rate < 25

  9. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  10. Characteristics and economy of the European reactor of pressurized water (EPR)

    International Nuclear Information System (INIS)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

    2005-01-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  11. Franco-German cooperation for the physical protection of the EPR reactor

    International Nuclear Information System (INIS)

    Jalouneix, J.; Hagemann, A.

    2001-01-01

    This article presents the proceeding that has been followed in the EPR (European pressurized water reactor) project concerning physical protection against malevolent actions and robbery of nuclear materials. Before the different options of the nuclear island were definitely set, a task group had been constituted to examine if these options could hamper the setting of physical protection measures that are required by the legislation of the 2 countries. Another group composed of experts from IPSN/GRS (Institut de Protection et de Surete Nucleaire / Gesellschaft fur Anlagen und Reaktorsicherheit) had the task to define common requirements concerning the physical protection of reactors in Germany and in France. In this framework the EPR project team has prepared a technical document reviewing the different dispositions that have been retained to assure the physical protection of the reactor. (A.C.)

  12. Soil class around me Serpong experimental power reactor (EPR) site plan base on micro tremor data

    International Nuclear Information System (INIS)

    Marjiyono; Soehaimi A; Hadi Suntoko; Yuliastuti; Syaeful H

    2015-01-01

    Surface geological characteristics has an important role on site response analysis in a region. In regard with experimental power reactor (EPR) construction plan in Serpong, the subsurface modeling from combination array and single station micro tremor data was done. The array and single station micro tremor measurement were performed in 9 and 90 sites, respectively, at ± 1 km radius around the EPR site plan. The Vs30 value was calculated from shear wave velocity structure around the investigated area. The soil classification based on Vs30 in the investigated area generally consists of SD (medium soil) and SC (soft rock) class. The EPR site plan its self in the SD class region. (author)

  13. The European Pressurized Water Reactor (EPR). State of the art after the preliminary design phase

    International Nuclear Information System (INIS)

    Bouteille, F.; Schneider, D.

    2002-01-01

    The European Pressurized Water Reactor (EPR) is an evolutionary development of the pressurized water reactor product lines built by Framatome and Siemens in France and Germany. Under the technical leadership of both nuclear power plant suppliers (now merged in Framatome ANP, a joint venture of AREVA and Siemens) the future-oriented plant concepts was developed in close cooperation with German and French utilities and in compliance with the European Utility Requirements. The EPR has safety features with which even extremely improbable, beyond design-basis events can be controlled and their effects can be limited to such an extent that no emergency response actions need be taken outside of the immediate plant site. This also means that safety systems prevent containment failure even in the improbable case of a core melt. This was confirmed by the French and German reactor safety authorities. The selected high thermal output also insures the economic viability of the innovative reactor concept, so that the power generation costs which can be achieved with the EPR will be absolutely competitive with those of fossil energy carriers. Framatome ANP has thus developed a pressurized water reactor ready for offer at the right time, which can completely fulfill the most rigorous requirements in terms of nuclear safety and economy. (Author)

  14. Finnish EPR Olkiluoto 3. The world's first third-generation reactor now under construction

    International Nuclear Information System (INIS)

    2007-01-01

    The EPR was developed by Framatome and Siemens KWU (the nuclear division of Siemens), whose nuclear activities were combined in January 2001 to form Framatome ANP, now AREVA NP. The French electricity utility EDF (Electricite de France), together with the major German utilities, played an active role in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. On December 18, 2003, the consortium formed by AREVA and Siemens - and led by AREVA - signed a contract with TVO for the turnkey construction of the EPR. The overall Olkiluoto 3 project cost has been estimated by TVO at around euros 3 Billion. TVO is responsible for the overall project management and licensing process with the Finnish Safety Authority STUK. In the pre-qualification phase, STUK concluded that the EPR can meet the Finnish licensing requirements. All specific comments will be taken into account for the realization of the project. In January 2005, STUK emphasized in its safety assessment that the evolutionary EPR design compared to predecessor product lines has been further enhanced by AREVA. This paper presents first, The Finnish energy situation (Electricity consumption and supply, Finland's Kyoto CO 2 cutback, Competitiveness of nuclear power), and then the EPR in Olkiluoto (General schedule of responsibilities, Important milestones of the project). Finally, the EPR third-generation and advanced reactor is presented with its position in the international competition (Targeted design objectives, Main characteristics, competitiveness, safety, Additional measures to prevent the occurrence of events likely to damage the core, Increased protection against the consequences of core melt)

  15. Energies and media nr 31. The EPR. Its role in the nuclear sector. Finland, Flamanville, Abu Dhabi. The reactor range

    International Nuclear Information System (INIS)

    2010-02-01

    After some comments on recent events in the nuclear sector in different countries (energy policy and projects in the USA, China, Italy, UK, Germany), this publication discusses the role of the EPR. It briefly outlines the characteristics of the third generation reactors compared with that of the first and second ones, evokes the influence of September 11 on design specifications, and evokes the international discussions about the project of fourth-generation reactors and the researches on nuclear fusion. It outlines the current context and the role of nuclear energy in the reduction of greenhouse gas emissions, briefly describes the opportunities offered by the use of thorium, and by fast neutron reactors. It comments the construction of the EPRs in Finland and in Flamanville, some characteristics of the EPR control system, and how France failed in selling the EPR to Abu Dhabi. It finally evokes the French offer in terms of nuclear reactors

  16. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  17. Tank type LMFBR type reactors

    International Nuclear Information System (INIS)

    Shimizu, Hiroshi

    1985-01-01

    Purpose: To detect the abnormality in the suspended body or reactor core supporting structures thereby improve the safety and reliability of tank type LMFBR reactors. Constitution: Upon inspection during reactor operation period, the top end of the gripper sensing rod of a fuel exchanger is abutted against a supporting bed and the position of the reactor core supporting structures from the roof slab is measured by a stroke measuring device. Then, the sensing rod is pulled upwardly to abut against the arm portion and the position is measured by the stroke measuring device. The measuring procedures are carried out for all of the sensing rods and the measured values are compared with a previously determined value at the initial stage of the reactor operation. As a result, it is possible to detect excess distortions and abnormal deformation in the suspended body or reactor core supporting structures. Furthermore, integrity of the suspended body against thermal stresses can be secured by always measuring the coolant liquid level by the level measuring sensor. (Kamimura, M.)

  18. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  19. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  20. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  1. The Flamanville EPR: an EDF third generation reactor to prepare the future

    International Nuclear Information System (INIS)

    2011-11-01

    This document presents the EPR reactor as an evolution of the currently operated nuclear reactors, but with a reinforced safety. It incorporates technological improvements which enable the number of incidents, the core fusion risk, the consequences of an accident, and workers' exposure to be reduced, the releases and the quantities of radioactive wastes to be limited, and the ecosystem to be preserved. It indicates the planning from public debate in 2004 to electricity production in 2016. It outlines the industrial cooperation for the plant construction, and economic consequences for the area

  2. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages

    International Nuclear Information System (INIS)

    Jurado P, M.; Martin del Campo M, C.

    2005-01-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  3. Position adopted by the government about the safety options of the EPR reactor project; Prise de position du gouvernement concernant les options de surete du projet de reacteur EPR

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-10-01

    On September 28, 2004, on behalf of the French ministers in charge of nuclear safety, the general director of nuclear safety and radiation protection addressed to the president of Electricite de France (EdF) a letter presenting the government's position about the safety options of the EPR (European Pressurized Reactor) project. On the basis of the examination carried out by the nuclear safety authority (ASN) and by the permanent group of reactor experts, the government has considered these options as satisfactory with respect to the safety improvement objectives. Therefore, the government requested EdF to comply with these technical rules for any future reactor development. This dossier includes: the letter of the government, the technical directives for the design and construction of the next generation of PWR-type reactors, the technical rules relative to the design of the main primary and secondary coolant circuits of PWR-type reactors, and the technical file about the safety of the EPR project reprinted from the 2003 report of nuclear safety and radiation protection authority. (J.S.)

  4. Nuclear accident: the EPR, the most dangerous reactor in the world

    International Nuclear Information System (INIS)

    Large, John

    2007-02-01

    After a brief description of the operation of a nuclear reactor, this document outlines the complexity of this machine and the problems which may happen and result in radioactive releases. It also briefly recalls past accidents (Chernobyl, Three Mile Island) and serious incidents (Fosmark in Sweden, Blayais in France). The author then describes how the nuclear industry and nuclear safety authorities tend to deny the risk, notably as far as the EPR is concerned. He outlines that the hypotheses retained by EDF and safety authorities for the quantity of radioactivity released in case of accident might be under-assessed as it appears when they are compared with data obtained by different models and in reality. The author considers that the use of MOX introduces difficulties to ensure nuclear safety, and that the EPR will therefore be the most dangerous nuclear reactor in the world. Based on a simulation, he shows that, in case of accident in of the EPR in Flamanville, radioactive fallouts will occur at the continental scale in less than two days. He proposes an assessment of consequences for the population in terms of deaths, cancers, evacuation. He indicates the different exposure modes for the population, and the countermeasures to be implemented in case of accident (he proposes an assessment of consequences in terms of morbidity with or without countermeasures)

  5. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  6. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  7. An Overview of Safety Barrier at EPR Reactor in France

    International Nuclear Information System (INIS)

    Mohd Zaid Mohamed; Syahirah Abdul Rahman; Mohamad Azman Che Mat Isa

    2011-01-01

    The fission of atomic nuclei, performed in reactors to generate heat, brings into play large quantities of radiation-emitting radioactive substances from which people and the environment must be protected. This paper explains the need for nuclear safety, which consists of the set of technical and organizational provisions taken at each stage in the design, construction and operation of a nuclear plant to ensure normal service, prevent the risks of an accident and limit its consequences in the unlikely event of its occurrence. (author)

  8. LMFBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  9. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  10. EPR compared to international requirements (Mainly EUR)

    International Nuclear Information System (INIS)

    Broecker, B.

    1996-01-01

    A number of European Utilities have entered an agreement to write common requirements dedicated to future light water nuclear power plants to be built in Europe. The activities are known under the sign EUR (European Utilities Requirements). EPR, the future European Pressurized water Reactor, is the first installation of this type which will be operational from the year 2000 onwards, must fulfill the European requirements. EPR will serve as a test whether these requirements are realistic and well balanced. At the basic design stage of EPR, this paper concentrates on four main topics: the requirements which are new compared with existing reactors and which put a major challenge to the designer; the requirements today still open and the way they can be met by the EPR or not; the points for which already today the EPR special requirements exceed the EUR; the examples where the design of the EPR has given feedback which has led to a change of the EUR. EPR and EUR are different approaches to the reactor of the future. EUR is a set of requirements which leaves a flexibility to the designer while EPR is a real project which defines the technical solutions. EPR will fulfill the EUR and will at the same time serve as a test whether these requirements are realistic. EPR will also fulfill international requirements with minor changes. (J.S.). 7 figs

  11. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  12. Regularized tripartite continuous variable EPR-type states with Wigner functions and CHSH violations

    International Nuclear Information System (INIS)

    Jacobsen, Sol H; Jarvis, P D

    2008-01-01

    We consider tripartite entangled states for continuous variable systems of EPR type, which generalize the famous bipartite CV EPR states (eigenvectors of conjugate choices X 1 - X 2 , P 1 + P 2 , of the systems' relative position and total momentum variables). We give the regularized forms of such tripartite EPR states in second-quantized formulation, and derive their Wigner functions. This is directly compared with the established NOPA-like states from quantum optics. Whereas the multipartite entangled states of NOPA type have singular Wigner functions in the limit of large squeezing, r → ∞, or tanh r → 1 - (approaching the EPR states in the bipartite case), our regularized tripartite EPR states show singular behaviour not only in the approach to the EPR-type region (s → 1 in our notation), but also for an additional, auxiliary regime of the regulator (s→√2). While the s → 1 limit pertains to tripartite CV states with singular eigenstates of the relative coordinates and remaining squeezed in the total momentum, the (s→√2) limit yields singular eigenstates of the total momentum, but squeezed in the relative coordinates. Regarded as expectation values of displaced parity measurements, the tripartite Wigner functions provide the ingredients for generalized CHSH inequalities. Violations of the tripartite CHSH bound (B 3 ≤ 2) are established, with B 3 ≅2.09 in the canonical regime (s → 1 + ), as well as B 3 ≅2.32 in the auxiliary regime (s→√2 + )

  13. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  14. FBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Tsugio.

    1986-01-01

    Purpose: To ensure the thermal integrity of a reactor vessel in FBR type reactors by preventing sodium vapors or the likes from intruding into a shielding chamber and avoiding spontaneous convection thereof. Constitution: There are provided a shielding plug for shielding the upper opening of a reactor container, an annular thermal member disposed to the circumferential side in the container, a shielding member for shielding upper end of the shielding chamber and a plurality of convection preventive plates suspended from the thermal member into the shielding chamber, and the shielding chamber is communicated by way of the relatively low temperature portion of the container with a gas communication pipe. That is, by closing the upper end of the shielding chamber with the shielding member, coolant vapors, etc. can be prevented from intruding into the shielding chamber. Further, the convection preventive plates prevent the occurrence of spontaneous convection in the shielding chamber. Further, the gas communication pipe absorbs the expansion and contraction of gases in the shielding chamber to effectively prevent the deformation or the like for each of the structural materials. In this way, the thermal integrity of the reactor container can surely be maintained. (Horiuchi, T.)

  15. Loop type LMFBR reactor

    International Nuclear Information System (INIS)

    Ito, Hiroyuki

    1989-01-01

    In conventional FBR type reactors, primary coolants at high temperature uprise at a great flow rate and, due to the dynamic pressure thereof, the free surface is raised or sodium is partially jetted upwardly and then fallen again. Then, a wave killing plate comprising a buffer plate and a deflection plate is disposed to the liquid surface of coolants. Most of primary sodium uprising from the reactor core along the side of the upper mechanism during operation collide against the buffer plate of the wave killing plate to moderate the dynamic pressure and, further, disperse radially of the reactor vessel. On the other hand, primary sodium passing through flowing apertures collides against the deflection plate opposed to the flowing apertures to moderate the dynamic pressure, by which the force of raising the free surface is reduced. Thus, uprising and waving of the free surface can effectively be suppressed to reduce the incorporation of cover gases into the primary sodium, so that it is possible to prevent in injury of the recycling pump, abrupt increase of the reactor core reactivity and reduction of the heat efficiency of intermediate heat exchangers. (N.H.)

  16. The aging of equipment in service: Preventive measures taken in the design of the EPR reactor

    International Nuclear Information System (INIS)

    Lefrancois, A.

    2009-01-01

    The design of a new Pressurized Water Reactor (PWR) has to take into account past experience, and in particular the lessons drawn from aging of components already in operation. The benefit of this process for tong term operation can be exceptionally high in the case of an evolutionary design such as that of the EPR TM reactor, for which the materials and the manufacturing conditions are the result of a very Long term optimization. Aging of materials in operation encompasses several damages, from the embrittlement caused by irradiation or thermal aging to the generation of cracks by fatigue or corrosion. These damages are known and characterized through international field experience and through a Cot of Laboratory studies performed worldwide, in France mainly by EDF, Cea and AREVA NP. Even if some damage mechanisms are not fully understood, good practice rules or empirical models were developed in order to give accurate predictions of damage kinetics and consequences. All this experience was taken into account during the design phase of the EPR TM project to define mate-rials and/or service conditions allowing to prevent or to minimize aging. (author)

  17. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  18. FBR type reactor

    International Nuclear Information System (INIS)

    Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

    1996-01-01

    In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

  19. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  20. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  1. 76 FR 26775 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on EPR; Cancellation to...

    Science.gov (United States)

    2011-05-09

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on EPR; Cancellation to May 11, 2011, ACRS Meeting-- Federal Register Notice The Federal Register Notice for the ACRS Subcommittee Meeting on the design certification application review of the U.S...

  2. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  3. Tank type nuclear reactors

    International Nuclear Information System (INIS)

    Naito, Kesahiro; Shimoyashiki, Shigehiro; Yokota, Norikatsu; Takahashi, Kazuo.

    1985-01-01

    Purpose: To improve the seismic proofness and the radiation shielding of LMFBR type reactors by providing the reactor with a structure reduced in the size and the weight, excellent in satisfactory heat insulating property and having radioactive material capturing performance. Constitution: Two sheets of ceramic plate members (for instance, mullite, steatite, beryllium ceramics or the like) which can be fabricated into plate-like shape and have high heat insulating property are overlapped with each other, between which magnetic heat-insulating material with magnetizing magnetic ceramics (for example, Lisub(0.5)Fesub(2.5)O 4 , Ni-Fe 2 O 4 , Fe-Fe 2 O 4 ) are sandwiched and the whole assembly is covered with metal coating material (for example, stainless steels). The inside of the coating material is evacuated or filled with an inert gas with low heat-conductivity (argon) at a pressure less than 1 kg/cm 2 abs, considering that the temperature goes higher and the inner pressure increases upon operation. In this way, the size of the laminated structure can be reduced to about 1/7 of the conventional case. The magnetic heat insulating materials can capture the magnetic impurities in sodium. (Kawakami, Y.)

  4. Economical opportunities on advanced conventional island design for the European pressurized water reactor (EPR) based on Konvoi design. Annex 6

    International Nuclear Information System (INIS)

    Kremayr, A.; Wagner, K.; Schuberth, U.

    2002-01-01

    Design of the European Pressurized Water Reactor (EPR) has been finalized by the end of 1998. In parallel with these efforts, the German utilities group contracted the Siemens AG Power generation Group (KWU) to develop an advanced and optimized conventional island for the EPR. The main objectives for improving the conventional island design were determined on the basis of experience of the Konvoi series plants and advanced fossil plants. This paper describes the innovations introduced to the conventional island and presents the reasons for the resultant cost reductions. (author)

  5. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  6. The nuclear renaissance and AREVA's reactor designs for the 21st century. EPR and SWR-1000

    International Nuclear Information System (INIS)

    Stosic, Z.V.

    2007-01-01

    Hydro and nuclear energy are the most environmentally benign way of producing electricity on a large scale. Nuclear generated electricity releases 38 times fewer greenhouse gases than coal, 27 times fewer than oil and 15 times fewer than natural gas [9]. On a global scale nuclear power annually saves about 10% of the global CO 2 emission. European nuclear power plants save amount of CO 2 emissions corresponding with the annual emission of CO 2 from all European passenger cars [16]. Also, that is approximately twice the total estimated quantity to be avoided in Europe under the Kyoto Protocol during the period 2008-2012. In respect to main drivers - such as concerns of the global warming effect, population growth, and future energy supply shortfall, low operating costs, reduced dependence on imported gas - it is clear that 30 new nuclear reactors currently being constructed in 11 countries and another 35 and more planed during next 10 years confirm the nuclear renaissance. Participation in the construction of 100 reactors out of 443 worldwide operated in January 2006 and supplying fuel to 148 of them AREVA helps meet the 21 st century's greatest challenges: making energy available to all, protecting the planet, and acting responsibly towards future generations. With EPR and SWR-1000, AREVA NP has developed advanced design concepts of Generation III+ nuclear reactors which fully meet the most stringent requirements in terms of nuclear safety, operational reliability and economic performance. (author)

  7. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    Energy Technology Data Exchange (ETDEWEB)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A. [AREVA - Tour AREVA, 1 Place Jean Millier, 92084 Paris La Defense (France)

    2013-07-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

  8. MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

    International Nuclear Information System (INIS)

    Arslan, M.; Villele, E. de; Gauthier, J.C.; Marincic, A.

    2013-01-01

    In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO 2 fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory

  9. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  10. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  11. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  12. Fusion reactors - types - problems

    International Nuclear Information System (INIS)

    Schmitter, K.H.

    1979-07-01

    A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

  13. Pool-type reactor

    International Nuclear Information System (INIS)

    Hopkins, S.R.

    1977-01-01

    This invention relates to a pool nuclear reactor fitted with a perfected system to raise the buckets into a vertical position at the bottom of a channel. This reactor has an inclined channel to guide a bucket containing a fuel assembly to introduce it into the reactor jacket or extract it therefrom and a damper at the bottom of the channel to stop the drop of the bucket. An upright vertically movable rod has a horizontally articulated arm with a hook. This can pivot to touch a radial lug on the bucket and pivot the bucket around its base in a vertical position, when the rod moves up [fr

  14. 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).

  15. FBR type reactor

    International Nuclear Information System (INIS)

    Yamaoka, Mitsuaki

    1988-01-01

    Purpose: To enable to increase the burning period by enabling to decrease the reduction of burning reactivity and unifying the irradiation amount of fast neutrons. Constitution: A cylindrical reactor core made of fissile material-enriched fuel is constituted so as to form a plurality of layer-like enriched regions in which the enrichment degree of the fissile material is increased from the center to the radial and axial directions. Then, the ratio between the average enrichment degree for all of the enrichment regions other than the region at the reactor core center with the lowest enrichment degree and the enrichment degree of the enriched region formed at the center of the reactor core is made greater by 5 % or 20 % than the ratio at the initial burning stage where the power distribution of the reactor core is most flattened. (Kawakami, Y.)

  16. BWR type reactors

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka

    1983-01-01

    Purpose: To decrease the control rod exchanging frequency by increasing the working life of control rods for ordinary operation with large neutron irradiation dose, to thereby decrease the exposure dose for operators performing exchanging work, as well as decrease the amount of radioactive wastes resulted upon exchange of the control rods. Constitution: Hafnium solid metal is employed as the neutron absorber of control rods for usual operation inserted into and withdrawn from fuel assemblies for the reactor power control over the entire cycle of the ordinary reactor operation and boron carbide powder is employed as the neutron absorber for emergency control rods to be inserted between the fuel assemblies only upon reactor scram or shutdown, whereby the working life of the control rods for ordinary reactor operation with greater neutron irradiation dose can be improved. Accordingly, the control rod exchanging frequency can be reduced to decrease the exposure dose to the operator for conducting the exchanging work. (Yoshihara, H.)

  17. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  18. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  19. BWR type reactors

    International Nuclear Information System (INIS)

    Hayashi, Katsuhisa; Watanabe, Shigeru.

    1983-01-01

    Purpose: To simplify the structure of control rod driving systems, as well as improve the safety and maintainability thereof. Constitution: Control-rod-guide tubes are disposed vertically above the reactor core and control-rod drives are disposed further thereabove, by which the control rods are moved upwardly and downwardly from above the reactor core through the guide tubes. Further, a partitioning cylinder is provided between the inner cirumferential wall at the upper portion of a pressure vessel and the control-rod-guide tubes and a gas-liquid separator is disposed to the space between the partitioning cylinder and the pressure vessel wall, to which steams generated in the reactor core are introduced. In such a structure of the reactor, since all of the control rods are inserted or extracted by the control rod drive system from above the reactor core, if the control rod drives or the likes should fail and accidentally drop the control rods, they exert in the direction of suppressing the nuclear reaction, whereby the safety can be improved. (Sekiya, K.)

  20. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  1. FBR type reactor

    International Nuclear Information System (INIS)

    Inoue, Kotaro; Kawashima, Katsuyuki; Zuketen, Atsushi.

    1982-01-01

    Purpose: To flatten the power distribution of a reactor core and shorten the breeding time. Constitution: The reactor core comprises a core region having fission products, an outer blanket region surrounding the outer side of the core region and having fertile material and an inner blanket region disposed within the core region and having fertile material. The axial thickness of the inner blanket region is made greater at the central portion and smaller at the peripheral portion of the inner blanket region, and the outermost peripheral end at the peripheral portion of the inner blanket region is opposed by way of the core region to the outer blanket region. In such an arrangement, the power decrease in the peripheral portion of the core region can be suppressed to thereby flatten the power distribution in the reactor core and shorten the breeding time. (Moriyama, K.)

  2. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  3. Natural convection type reactor

    International Nuclear Information System (INIS)

    Nakayama, Takafumi; Horiuchi, Tetsuo; Moriya, Kimiaki; Matsumoto, Masayoshi; Akita, Minoru.

    1988-01-01

    Purpose: To improve the reliability by decreasing the number of dynamic equipments and safely shutdown the reactor core upon occurrence of accidents. Constitution: A pressure relief valve and a pressurizing tank or gravitational water falling tank disposed to the main steam pipe of a reactor are installed in combination. Upon loss-of-coolant accident, the pressure relief valve is opened to reduce the pressure in the reactor pressure vessel to the operation pressure for each of the tanks, thereby enabling to inject water in the pressurizing tank at first and, thereafter, water in the gravitational water falling tank successively to the inside of the pressure vessel. By utilizing the natural force in this way, the reliability can be improved as compared with the case of pumped water injection. Further, by injecting an aqueous boric acid to a portion of a plurality of tanks, if the control rod insertion becomes impossible, aqueous boric acid can be injected. (Takahashi, M.)

  4. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    International Nuclear Information System (INIS)

    Arkoma, Asko; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-01-01

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  5. Statistical analysis of fuel failures in large break loss-of-coolant accident (LBLOCA) in EPR type nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Arkoma, Asko, E-mail: asko.arkoma@vtt.fi; Hänninen, Markku; Rantamäki, Karin; Kurki, Joona; Hämäläinen, Anitta

    2015-04-15

    Highlights: • The number of failing fuel rods in a LB-LOCA in an EPR is evaluated. • 59 scenarios are simulated with the system code APROS. • 1000 rods per scenario are simulated with the fuel performance code FRAPTRAN-GENFLO. • All the rods in the reactor are simulated in the worst scenario. • Results suggest that the regulations set by the Finnish safety authority are met. - Abstract: In this paper, the number of failing fuel rods in a large break loss-of-coolant accident (LB-LOCA) in EPR-type nuclear power plant is evaluated using statistical methods. For this purpose, a statistical fuel failure analysis procedure has been developed. The developed method utilizes the results of nonparametric statistics, the Wilks’ formula in particular, and is based on the selection and variation of parameters that are important in accident conditions. The accident scenario is simulated with the coupled fuel performance – thermal hydraulics code FRAPTRAN-GENFLO using various parameter values and thermal hydraulic and power history boundary conditions between the simulations. The number of global scenarios is 59 (given by the Wilks’ formula), and 1000 rods are simulated in each scenario. The boundary conditions are obtained from a new statistical version of the system code APROS. As a result, in the worst global scenario, 1.2% of the simulated rods failed, and it can be concluded that the Finnish safety regulations are hereby met (max. 10% of the rods allowed to fail)

  6. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  7. FBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Azekura, Kazuo; Inoue, Kotaro.

    1981-01-01

    Purpose: To decrease power fluctuations due to burning of blanket fuel element clusters by partially replacing the fertile materials in the blanket fuel element clusters with fissile materials. Constitution: Fertile materials in the radial blanket fuel element clusters disposed to the outside or inside of the reactor core are partially replaced with fissile materials. Since the power density of the fissile materials is at the maximum in the initial burning stage and decreases as the burning proceeds, the power density of the materials which is smaller in the initial burning stage and becomes greater with the burning by the neutron-accumulated plutonium is offset. Accordingly, the power fluctuations in the blanket fuel element clusters due to the burning made smaller thereby enable to form a reactor core with less power fluctuations due to burning under the constant coolant flow rate depending on the power in the final burning stage where the blanket power is maximum. (Moriyama, K.)

  8. FBR type reactors

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

    1986-01-01

    Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

  9. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part 'CIRCUITS' regroups under a condensed form - in French and using international units - the essential information contained in both basic documents of the American project for a molten-salt breeder power plant. This part is only dealing with things relating to the CEA-EDF workshop 'CIRCUITS'. It is not concerned with information on: the reactor and the moderator replacement, the primary and secondary salts, and the fuel salt reprocessing, that are dealt with in parts 'CORE' and 'CHEMISTRY' respectively. The possible evolutions in the data - and solutions - taken by the American designers for their successive projects (1970 to 1972) are shown. The MSBR power plant comprises three successive heat transfer circuits. The primary circuit (Hastelloy N), radioactive and polluted, containing the fuel salt, includes the reactor, pumps and exchangers. The secondary circuit (pipings made of modified Hastelloy N) contaminated in the exchanger, ensures the separation between the fuel and the fluid operating the turbo-alternator. The water-steam circuit feeds the turbine with steam. This steam is produced in the steam generator flowed by the secondary fluid. Some subsidiary circuits (discharge and storage of the primary and secondary salts, ventilation of the primary circuit ...) complete the three principal circuits which are briefly described. All circuits are enclosed inside the controlled-atmosphere building of the nuclear boiler. This building also ensures the biological protection and the mechanical protection against outer aggressions [fr

  10. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. Emphasize is put essentially on the fuel salt of the primary circuit inside which fission reactions occur. The reasons why the (LiF-BeF 2 -ThF 4 -UF 4 ) salt was chosen for the M.S.B.R. concept are examined; the physical, physicochemical and chemical properties of this salt are discussed with its interactions with the structural materials and its evolution in time. An important part of this volume is devoted to the continuous reprocessing of the active salt, the project designers having deemed advisable to take advantage at best from the availability of a continuous purification, in a thermal breeding. The problem of tritium formation and distribution inside the reactor is also envisaged and the fundamentals of the chemistry of the secondary coolant salt are given. The solutions proposed are: the hydrogen scavenging of the primary circuit, a reduction in metal permeability by an oxyde layer deposition on the side in contact with the vapor, and tritium absorption through an isotope exchange with the hydroxifluoroborate [fr

  11. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  12. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  13. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  14. BWR type reactors

    International Nuclear Information System (INIS)

    Tsunoyama, Shigeaki; Tanabe, Akira.

    1979-01-01

    Purpose: To provide a main steam pressure shock absorber for reflecting the effect of the pressure propagation to coolants surface in the reactor core. Constitution: An annular shock absorber having near the water level through holes for water level measurement is provided to the gap between the skirt of a steam separator and a pressure vessel. Pressure waves are made the rapid closure of a main steam check valve. If arrived from the dome to the shock absorber, are mostly reflected to the side of the dome and give no substantial effects on the water surface. If the through holes are made small enough, the effects of pressure waves passing through the holes are negligible if they reach the water surface. (Kawakami, Y.)

  15. Molten salt reactor type

    International Nuclear Information System (INIS)

    1977-01-01

    This document is one of the three parts of a first volume devoted to the compilations of American data on the molten salt reactor concept. This part describes the MSBR core (data presented are from ORNL 4541). The principal characteristics of the core are presented in tables together with plane and elevation drawings, stress being put upon the reflector, and loading and unloading. Neutronic, and thermal and hydraulic characteristics (core and reflectors) are more detailed. The reasons why a graphite with a tight graphite layer has been chosen are briefly exposed. The physical properties of the standard graphite (irradiation behavior) have been determined for an isotropic graphite with fine granulometry; its dimensional variations largely ressemble that of Gilsonite. The mechanical stresses computed (Wigner effect) do not implicate in any way the graphite stack [fr

  16. LMFBR type reactors

    International Nuclear Information System (INIS)

    Sakurai, Akio; Matsushita, Kazuo.

    1985-01-01

    Purpose: To surely prevent the vibrational displacement of a rector core, and remove the sliding portions so as to avoid sticking or localized load concentration. Constitution: Cylindrical vibration-damping walls are protruded from the inner surface at the bottom of the main vessel. The vibration-damping walls constituted with outer and inner walls defining a coolant passage surround the reactor core support structures with small gaps between the inner surface of the inner wall and the outer circumferential surface of the core support structures. If the core support structures tend to displace horizontally due to earthquakes or the likes, the small gaps are varied and the coolants flow through the gap. Vibration-damping for the core support structures can be obtained by the flowing resistance to the coolants. (Yoshino, Y.)

  17. Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

    1975-01-01

    A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

  18. Modern light-water reactors - EPR and SWR 1000. Present status and possibilities of development and application

    International Nuclear Information System (INIS)

    Brettschuh, W.; Schneider, D.

    2001-01-01

    A number of reactor concepts are being advanced worldwide also in the light of long-term perspectives in the power economy. Activities in this respect are concentrated on the proven technologies of light-water and heavy-water reactors and on gas or liquid-metal cooled designs. Also smaller units are being considered for various applications. Framatome ANP focus on further development of the EPR and SWR 1000 light-water reactor lines. These plant concepts, which were developed in cooperation with, and with the support of, power utilities together with a number of European partners, are characterized by an even higher level of safety, especially as far as potential accidental releases are concerned, and by their competitiveness in the electricity generating market compared to other sources of energy. Given the cost structure of these plants, with approx. 60% capital costs, approx. 24% operating costs, and approx. 16% fuel cycle costs, efforts are made in particular to reduce the dominating capital cost share. For the EPR, this is to be achieved by high plant capacities, i.e. lower specific capital outlays per kilowatt installed; for the SWR 1000, mainly by means of an optimized, streamlined plant design. Moreover, installed nuclear generating capacities provide long-term assurance in calculating electricity generating costs, as they are less sensitive to the variable costs of the fuel cycle and of operation than fossil-fired plants. Commercial maturity of the SWR 1000 will be achieved after the completion of basic development in late 1999 and the ongoing detailed design steps in 2001. Also for the EPR, commercial maturity has been achieved with the completion of the basic design phase. (orig.) [de

  19. Investigations in quantum games using EPR-type set-ups

    Science.gov (United States)

    Iqbal, Azhar

    2006-04-01

    Research in quantum games has flourished during recent years. However, it seems that opinion remains divided about their true quantum character and content. For example, one argument says that quantum games are nothing but 'disguised' classical games and that to quantize a game is equivalent to replacing the original game by a different classical game. The present thesis contributes towards the ongoing debate about quantum nature of quantum games by developing two approaches addressing the related issues. Both approaches take Einstein-Podolsky-Rosen (EPR)-type experiments as the underlying physical set-ups to play two-player quantum games. In the first approach, the players' strategies are unit vectors in their respective planes, with the knowledge of coordinate axes being shared between them. Players perform measurements in an EPR-type setting and their payoffs are defined as functions of the correlations, i.e. without reference to classical or quantum mechanics. Classical bimatrix games are reproduced if the input states are classical and perfectly anti-correlated, as for a classical correlation game. However, for a quantum correlation game, with an entangled singlet state as input, qualitatively different solutions are obtained. The second approach uses the result that when the predictions of a Local Hidden Variable (LHV) model are made to violate the Bell inequalities the result is that some probability measures assume negative values. With the requirement that classical games result when the predictions of a LHV model do not violate the Bell inequalities, our analysis looks at the impact which the emergence of negative probabilities has on the solutions of two-player games which are physically implemented using the EPR-type experiments.

  20. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors; Actualizacion de los costos de combustible nuclear de la recarga de equilibrio de los reactores ABWR y EPR

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, R.F. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: rortega@fi-b.unam.mx

    2008-07-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U{sub 3O}8 in January, 2005 to a maximum of US$137.00 dollars by Ib U{sub 3}O{sub 8} by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U{sub 3}O{sub 8} such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  1. Updating of the costs of the nuclear fuels of the equilibrium reloading of the A BWR and EPR reactors

    International Nuclear Information System (INIS)

    Ortega C, R.F.

    2008-01-01

    In the last two and a half years, the price of the uranium in the market spot has ascended of US$20.00 dollars by lb U 3O 8 in January, 2005 to a maximum of US$137.00 dollars by Ib U 3 O 8 by the middle of 2007. At the moment this price has been stabilized in US$90.00 dollars by Ib U 3 O 8 such for the market spot, like for the long term contracts. In this work the reasons of this increment are analyzed, as well as their impact in the fuel prices of the balance recharge of the advanced reactors of boiling water (A BWR) and of the advanced water at pressure reactors (EPR). (Author)

  2. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  3. Expertise of safety of the Flamanville EPR reactor. Manche inhabitants worry, the Institute answer them

    International Nuclear Information System (INIS)

    2014-01-01

    A first article gives some examples of activities and of the influence of experts who, at each step of the construction of the EPR in Flamanville and until its operation, check whether safety requirements are met. For example, strain gauges are used to control the behaviour of the containment all along its construction. Safety objectives for the EPR have been defined by French and German authorities who decided some design evolutions: number of emergency generators, use of a new water resource to cool the fuel in case of accident, safety systems. In a brief interview, the EDF technical manager evokes the lessons learned from Fukushima from a technical point of view as well as in terms of personnel, and evokes some modifications of the EPR project. A brief article evokes the participation of the IRSN to the Finnish EPR project. A last article addresses the cooperative activities of the local information commissions and other involved actors (IRSN, ASN, EDF, Areva) about the safety of the installation

  4. The EPR (European Pressurized Water Reactor) containment - concept, testing of leakage behaviour, FRP liner

    Energy Technology Data Exchange (ETDEWEB)

    Touret, J.P. [EDF SEPTEN, Villeurbanne (France); Liersch, G. [Bayernwerk Kerenergie GmbH, Muenchen (Germany); Danisch, R. [Siemens AG, KWU NAD, Erlangen (Germany)

    2001-07-01

    The Basic Design of the EPR has now been completed. The containment plays a major safety-related role with respect to protection of the environment against radioactive releases. The EPR features a double (steel-reinforced concrete/prestressed concrete) containment design, with the inner containment coated additionally with a fibreglass-reinforced plastic (FRP) liner in certain areas. This means that containment leaktightness is provided mainly by the prestressed concrete and the FRP liner in the event of a postulated accident. The numerous findings of the tests carried out so far in both France and Germany are summarized. (orig.) [German] Das Basic Design fuer den EPR ist fertiggestellt. Entscheidend fuer eine Realisierung wird neben der politischen Akzeptanz vor allem die Wettbewerbsfaehigkeit mit anderen Energietraegern sein. Im EPR-Projekt wird der hohe Sicherheitsstandard der heutigen Kernkraftwerke in Deutschland und Frankreich ergaenzt, indem zusaetzlich technische Massnahmen ergriffen werden, um die Konsequenzen beim unterstellten Versagen aller sicherheitstechnischen Einrichtungen mit der Folge eines postulierten Niederschmelzen des Kerns technisch zu beherrschen. (orig.)

  5. Taking into account radiation protection for the EPR (European pressurized water reactor) design

    International Nuclear Information System (INIS)

    Michoux, X.

    2005-01-01

    For a designer, the taking into account of radiation protection for the EPR design is based on several thrusts which concern different scopes as choice of materials, checking of design's options, layout of components and systems able to contain radioactivity in different states of operation (i.e.: pressurizer, tanks, actives systems separated from non actives systems), or the optimization of shielding according to the estimated maintenance during outage or during power operation. The EPR method used for radiation protection studies is close to the safety method (use of dose gauge, demonstration of radiation protection, works with high stake regarding the radiation protection studied in priority, parametric studies with use of one field Radiation protection...). Results of this method place EPR in a satisfactory progress compared to the best existing nuclear plants, regarding collective doses and privileging the most exposed workers. This method has also induced on the EPR Project the choice of working during power operation in order to obtain shorts outages, scrupulously respecting security rules, radiation protection and human factor. (author)

  6. Results of the basic design of the European Pressurized Water Reactor (EPR) and further prospective

    International Nuclear Information System (INIS)

    Fischer, U.

    1998-01-01

    The EPR is an evolutionary design, with maximum consideration given to the experience accumulated by Framatome and Siemens with the existing designs of plants successfully operating.However, substantial improvements with regard to availability, safety and economy have been included. The main target of the basic design phase was a clear definition of technical features based on common technical codes that comply with economic targets and licensability requirements in France and Germany.The EPR basic design represents a new design for a nuclear power plant that complies with the latest laws and regulations, both in France and Germany as well as in most of the countries utilizing nuclear power. Cost estimates prove the design economically viable

  7. Note related to the examination of fitness to service of the vessel bottom and cover of the Flamanville EPR reactor. Pedagogical sheet: Analysis of consequences of the fabrication anomaly of the Flamanville EPR reactor vessel bottom and cover

    International Nuclear Information System (INIS)

    2017-01-01

    After having recalled the context related to the discovery of a fabrication anomaly in the chemical composition of the steel used in the central part of the vessel cover and bottom of the Flamanville EPR reactor, this report presents the approach proposed and adopted by Areva for the justification of the mechanical strength of the vessel cover and bottom. It indicates the main conclusions of the IRSN and ASN on various issues addressed by the audit performed by AREVA: fabrication controls, material characterisation, thermal-mechanical loadings, mechanical analysis of a risk of sudden failure. The next part proposes a pedagogical sheet of consequences of a fabrication anomaly: presentation of vessel structure and specificities, report of the discovery of this anomaly and main consequence (steel embrittlement), expertise approach, expertise conclusions (control of the absence of any deleterious defect, characterisation of material mechanical properties, assessment of thermal-mechanical loadings, assessment of a risk of sudden failure, and in-service follow-up)

  8. Comparison of EPR response of alanine and Gd₂O₃-alanine dosimeters exposed to TRIGA Mainz reactor.

    Science.gov (United States)

    Marrale, M; Schmitz, T; Gallo, S; Hampel, G; Longo, A; Panzeca, S; Tranchina, L

    2015-12-01

    In this work we report some preliminary results regarding the analysis of electron paramagnetic resonance (EPR) response of alanine pellets and alanine pellets added with gadolinium used for dosimetry at the TRIGA research reactor in Mainz, Germany. Two set-ups were evaluated: irradiation inside PMMA phantom and irradiation inside boric acid phantom. We observed that the presence of Gd2O3 inside alanine pellets increases the EPR signal by a factor of 3.45 and 1.24 in case of PMMA and boric acid phantoms, respectively. We can conclude that in the case of neutron beam with a predominant thermal neutron component the addition of gadolinium oxide can significantly improve neutron sensitivity of alanine pellets. Monte Carlo (MC) simulations of both response of alanine and Gd-added alanine pellets with FLUKA code were performed and a good agreement was achieved for pure alanine dosimeters. For Gd2O3-alanine deviations between MC simulations and experimental data were observed and discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

  9. OMR type process heat reactor

    International Nuclear Information System (INIS)

    Franzetti, Franco.

    1974-01-01

    A description is given of an OMR type reactor for heat generation. It includes a vessel the upper part of which is shut by a plug. The lower part of the vessel includes a core of fuel elements and is filled with an organic liquid. Over this there is a middle area filled with an inert gas. The plug includes an upper part forming a closure and resting around its edge on the vessel, and a lower part fixed under the closure and composed of a hollow cylindrical tank fitted with a bottom and filled with another organic liquid. The height of the cylindrical tank is such that, increased by the height of the first organic liquid in the lower area and above the core, it provides biological protection. The cooling system includes a heat exchanger and a pump to move the liquid from the lower part of the core and to inject some as spray into that part of the vessel filled with the inert gas. When loading and unloading, after the reactor is shut down, the clear organic liquid contained in the plug is discharged into the reactor vessel in such a way that it does not mix with the opaque organic liquid already contained in the vessel, and in that the opaque organic liquid is emptied out [fr

  10. Analyzing three-player quantum games in an EPR type setup.

    Directory of Open Access Journals (Sweden)

    James M Chappell

    Full Text Available We use the formalism of Clifford Geometric Algebra (GA to develop an analysis of quantum versions of three-player non-cooperative games. The quantum games we explore are played in an Einstein-Podolsky-Rosen (EPR type setting. In this setting, the players' strategy sets remain identical to the ones in the mixed-strategy version of the classical game that is obtained as a proper subset of the corresponding quantum game. Using GA we investigate the outcome of a realization of the game by players sharing GHZ state, W state, and a mixture of GHZ and W states. As a specific example, we study the game of three-player Prisoners' Dilemma.

  11. An EPR study of the lacquer-type bottle deposits from red wines

    International Nuclear Information System (INIS)

    Troup, G.J.; Hutton, D.R.; Pilbrow, J.R.; Hunter, C.A.; Hewitt, D.G.

    1996-01-01

    Full text: Lacquer-type deposits from 4 naturally aged red wines and one 'artificially', aged wine (by heating for 16 hrs at 84 deg C) were obtained from the Australian Wine Research Institute, Adelaide, and examined by EPR spectroscopy with a Varian E- 12 spectrometer working at ∼ 9.1 GHz The naturally aged wine deposits showed Fe 3+ in a low-symmetry site (g = 4.3), and Cu 2+ in an axial symmetry site (as deduced from the hyperfine structure) showing superhyperfine structure from N nuclei. A strong free radical signal was also present. The artificially aged wine deposit spectrum was similar, except no superhyperfine structure was present. The deposits are known to be anthocyanines cross-linked to a protein. The older the wine, the more intense was the superhyperfine structure, due to copper-amine interaction with the protein. The free radical signal is associated with the anthocyanines

  12. United Kingdom and USSR reactor types

    International Nuclear Information System (INIS)

    Lewins, Jeffery

    1988-01-01

    The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

  13. High conversion burner type reactor

    International Nuclear Information System (INIS)

    Higuchi, Shin-ichi; Kawashima, Masatoshi

    1987-01-01

    Purpose: To simply and easily dismantle and reassemble densified fuel assemblies taken out of a high conversion ratio area thereby improve the neutron and fuel economy. Constitution: The burner portion for the purpose of fuel combustion is divided into a first burner region in adjacent with the high conversion ratio area at the center of the reactor core, and a second burner region formed to the outer circumference thereof and two types of fuels are charged therein. Densified fuel assemblies charged in the high conversion ratio area are separatably formed as fuel assemblies for use in the two types of burners. In this way, dense fuel assembly is separated into two types of fuel assemblies for use in burner of different number and arranging density of fuel elements which can be directly charged to the burner portion and facilitate the dismantling and reassembling of the fuel assemblies. Further, since the two types of fuel assemblies are charged in the burner portion, utilization factor for the neutron fuels can be improved. (Kamimura, M.)

  14. Livermore pool-type reactor

    International Nuclear Information System (INIS)

    Mann, L.G.

    1977-01-01

    The Livermore Pool-Type Reactor (LPTR) has served a dual purpose since 1958--as an instrument for fundamental research and as a tool for measurement and calibration. Our early efforts centered on neutron-diffraction, fission, and capture gamma-ray studies. During the 1960's it was used for extensive calibration work associated with radiochemical and physical measurements on nuclear-explosive tests. Since 1970 the principal applications have been for trace-element measurements and radiation-damage studies. Today's research program is dominated by radiochemical studies of the shorter-lived fission products and by research on the mechanisms of radiation damage. Trace-element measurement for the National Uranium Resource Evaluation (NURE) program is the major measurement application today

  15. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  16. The construction of the first EPR (European Pressurized water Reactor) in France in Flamanville

    International Nuclear Information System (INIS)

    2009-01-01

    This document outlines the importance of nuclear energy for France which is a way to guarantee energy independence, energy security, to struggle against climate change, and to answer to fossil energy depletion. It presents the EDF's electricity production stock and outlines on the one hand the major role it has in the French energy mix, and on the other hand the investments it requires. It presents the EPR as an industrial challenge for the future (an essential step in preparing the electricity production stock renewal, a demonstration of an industrial know-how, a reinforced competitiveness, a high security level, a reinforced protection of the environment, enhanced performance). Then, it presents the construction project from August 2006, and outlines that this project will promote regional development. Some key figures are given in appendix

  17. Design improvements related to chemistry for the Evolutionary Power reactor (EPR) unit at Flamanville 3

    International Nuclear Information System (INIS)

    Lacroix, Christophe; Alves-Vieira, Maria; Jutel, Laurent

    2012-09-01

    A significant number of improvements, including for chemistry, have been made for the design of the EPR turbine hall as a result of the considerable wealth of experience and feedback drawn from the EDF fleet. To reduce issues associated with flow accelerated corrosion, as well as Steam Generator (SG) fouling or clogging, appropriate materials for the piping and the exchangers, and an adequate chemical conditioning, were chosen to ensure a 60 year lifespan of the main components of the circuit. The condenser technology was also improved to mitigate the risks of raw water ingress. A dedicated local sampling line was added at feedwater (FW) to monitor the iron concentration. EPR includes a polishing system at start-up to ensure the purification of feedwater in a short time (less than 16 hours) along with reduced discharged effluents. Comprehensive studies were carried out, taking into account costs (investment, operation, and maintenance), consumption and discharge of chemical reagents, or risks of pollutions during the process. This lead to the choice of system dedicated to start-up only, similar to the mobile system used in some French Nuclear Power Plants (NPPs), instead of a condensate polishing plant. Emphasis was also placed on providing a flexible and secure injection system for the chemical reagents (SIR). Indeed, it will be possible to inject two amines and hydrazine, and to perform tailored injection in many parts of the circuit. Furthermore, the room was design to facilitate future design changes, such as if dispersant injection is required as a preventative solution for SG fouling. In addition to that, close attention was applied to mitigate chemical hazards, especially regarding the hydrazine CMR effect or explosive atmosphere. This paper describes the technical and the economical choices that lead to these improvements related to chemistry in the turbine hall design, along with a comprehensive overview of the chosen designs. (authors)

  18. On reactor type comparisons for the next generation of reactors

    International Nuclear Information System (INIS)

    Alesso, H.P.; Majumdar, K.C.

    1991-01-01

    In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs

  19. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  20. Synthesis of the IRSN report on the control-command architecture of the Flamanville 3 EPR reactor and the associated platforms

    International Nuclear Information System (INIS)

    2009-06-01

    This document briefly reports the IRSN's comments on the robustness of the control-command architecture of the Flamanville 3 EPR (European pressurized reactor), that is of the equipment which, to various extent, contribute to the installation safety. This report identifies the equipment the robustness of which should be improved, outlines the problems created by the complexity of computerized systems. Notably, the authors comment the ability of the off-the-shelf computer platforms acquired by EDF to deal with the nuclear context

  1. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    Energy Technology Data Exchange (ETDEWEB)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

    2008-07-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  2. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    International Nuclear Information System (INIS)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

    2008-01-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  3. Severe accident mitigation and core melt retention in the European pressurized reactor (EPR)

    International Nuclear Information System (INIS)

    Fischer, Manfred

    2003-01-01

    For the mitigation of severe accidents, the FPR has adopted and improved the defense-in-depth approaches of its predecessors, the French 'N4' and the German 'Konvoi' PWR's. Beyond these evolutionary changes, it includes a new, 4-th level of defense aimed at limiting the consequences of a postulated severe accident with core melting. This involves a strengthening of the confinement function and the avoidance of large early releases, by the prevention of scenarios and events with potentially high loads on the containment, incl. RPV failure at high pressure. The remaining low-pressure accidents are mitigated by dedicated design measures. The paper gives an overview and of the measures for H 2 -mitigation and steam explosion and focuses on a detailed description of the precautions and design measures for the stabilization and long-term cooling of the molten core. In the EPR the latter is achieved by melt spreading into a large outside-cooled crucible lateral to the pit, which is passively flooded and cooled with water from the IRWST. The separation of functions between pit and spreading room not only isolates the core catcher from the various loads during RPV failure, but also avoids any risks related to an unintended initiation of flooding during power operation. A stable state of the melt is reached after a few hours. Complete solidification is achieved within days. The core catcher can optionally be cooled actively by the CHRS, which avoids further steaming into the containment and establishes ambient pressure conditions in the long term. (author)

  4. Reactor types for the future

    International Nuclear Information System (INIS)

    Hall, A.C.

    1990-01-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  5. Reactor types for the future

    Energy Technology Data Exchange (ETDEWEB)

    Hall, A C [PWR Power Projects Ltd., Knutsford, Cheshire (United Kingdom)

    1990-06-01

    The factors impacting a utility's choice of reactor for commercial exploitation are discussed. Concepts available in time frames of 5, 10 and 20 years are considered. It is concluded that future programmes are likely to be based on a relatively small number of largely pre-licensed turnkey station designs. The near future is likely to be dominated by light water reactors. The Westinghouse AP600 design is briefly described. (author)

  6. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  7. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  8. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.

    2008-01-01

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  9. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  10. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    Ignatiev, V.; Devell, L.

    1995-01-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  11. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  12. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

  13. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  14. Loop-type FBR reactor

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

    1996-01-01

    The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

  15. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1975-01-01

    Object: To permit safe and reliable replacement of primary pipes by providing a reactor container so as to surround a pressure pipe, with upper portions of the two separably coupled together, and coupling the pressure pipe and primary piping by joint coupling above and below the reactor container, with the lower coupling joint surrounded by drain receptacle. Structure: At the time of replacement of a pressure pipe, a partition valve is opened to exhaust primary cooling water within pressure pipe and upper and lower portions of the primary piping and replace the decelerator within the reactor container with water of the same quality as that of pool water within an upper shield pool. Thereafter, the entire space above the drain receptacle is filled with pool water by closing a partition valve and opening a water supply valve. Then, upper portion seal cover, pool bottom lid, upper joint and upper portion primary piping are removed, then bolts and nuts are loosened, and the pressure pipe is taken out together with the shield block. (Kamimura, M.)

  16. Strategies of development of reactor types

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    The development of nuclear energy in the coming decades will depend on the goals followed, on the available technologies and on the strategies implemented in the world in agreement with public acceptation. This article is limited to the technical aspects of the strategies of development of reactor types: 1 - objectives; 2 - common constraints to all reactor types: safety and terrorism risks, wastes, non-proliferation, economics; 3 - different reactor types: general considerations, proven technologies (PWR, BWR, Candu), non-proven technologies but having an important experience, technologies at the design stage; 4 - energy systems and 'Generation IV forum': systems based on thermal neutron reactors and low enrichment, systems for the valorization of 238 U, systems for Pu burning, systems allowing the destruction of minor actinides, thorium-based systems, the Gen IV international forum; 5 - conclusion. (J.S.)

  17. EPR: Evidence and fallacy.

    Science.gov (United States)

    Nichols, Joseph W; Bae, You Han

    2014-09-28

    The enhanced permeability and retention (EPR) of nanoparticles in tumors has long stood as one of the fundamental principles of cancer drug delivery, holding the promise of safe, simple and effective therapy. By allowing particles preferential access to tumors by virtue of size and longevity in circulation, EPR provided a neat rationale for the trend toward nano-sized drug carriers. Following the discovery of the phenomenon by Maeda in the mid-1980s, this rationale appeared to be well justified by the flood of evidence from preclinical studies and by the clinical success of Doxil. Clinical outcomes from nano-sized drug delivery systems, however, have indicated that EPR is not as reliable as previously thought. Drug carriers generally fail to provide superior efficacy to free drug systems when tested in clinical trials. A closer look reveals that EPR-dependent drug delivery is complicated by high tumor interstitial fluid pressure (IFP), irregular vascular distribution, and poor blood flow inside tumors. Furthermore, the animal tumor models used to study EPR differ from clinical tumors in several key aspects that seem to make EPR more pronounced than in human patients. On the basis of this evidence, we believe that EPR should only be invoked on a case-by-case basis, when clinical evidence suggests the tumor type is susceptible. Copyright © 2014 Elsevier B.V. All rights reserved.

  18. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  19. Fuel exchanger in FBR type reactor

    International Nuclear Information System (INIS)

    Shinden, Kazuhiko; Tanaka, Osamu.

    1990-01-01

    The present invention concerns a fuel exchanger for exchanging fuels in an LMFBR type reactor using liquid metals as coolants. An outer gripper cylinder rotating device for rotating an outer gripper cylinder that holds a gripper is driven, to lower the gripper driving portion and the outer gripper cylinder, fuels are caught by the finger at the top end of the outer gripper cylinder and elevated to extract the fuels from the reactor core. Then, the gripper driving portion casing and the outer gripper cylinder are rotated to rotate the fuels caught by the gripper. Subsequently, the gripper driving portion and the outer gripper cylinder are lowered to charge the fuels in the reactor core. This can directly shuffle the fuels in the reactor core without once transferring the fuels into a reactor storing pot and replacing with other fuels, thereby shortening the shuffling time. (I.N.)

  20. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  1. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  2. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  3. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

  4. Radioactive waste management at WWER type reactors

    International Nuclear Information System (INIS)

    1993-05-01

    This report was prepared within the framework of the Technical Assistance Regional Project on Advice on Waste Management at WWER Type Reactors, which was initiated by the IAEA in 1991. The Regional Project is an integral part of the IAEA's activities directed towards improvement of the safety and reliability of nuclear power plants with WWER type reactors (Soviet designed PWRs). Forty-five WWER type units are currently in operation and twenty-five are under construction in Bulgaria, Czechoslovakia, Finland, Hungary and the former USSR. The idea of regional collaboration between eastern European countries under the auspices of the IAEA was discussed for the first time during the last meeting of the Council for Mutual Economic Assistance (CMEA) on spent fuel and radioactive waste management, held in Rez, Czechoslovakia, in October 1990. Since then, the CMEA and some of its former Member States have ceased to exist. However, there are many reasons for eastern European countries to continue their regional collaboration at a higher level. The USSR, the designer and supplier of WWER type reactors in eastern European countries, participated in the first phase of the project. The majority of WWER type reactors are situated in States of the former USSR (Russia and Ukraine). The main results of the first phase of the Regional Project are: (i) Re-establishment of communication channels among eastern European countries operating WWER type reactors by incorporating the IAEA's technical assistance; (ii) Identification of common waste management problems (administrative and technical) requiring resolution; (iii) Familiarization with radioactive waste management systems at nuclear power plants with WWER type reactors - Paks (Hungary), Loviisa (Finland), Jaslovske Bohunice (Czechoslovakia) and Novovoronezh (Russian Federation). Tabs

  5. Vibration-proof FBR type reactor

    International Nuclear Information System (INIS)

    Kawamura, Yutaka.

    1992-01-01

    In a reactor container in an FBR type reactor, an outer building and upper and lower portions of a reactor container are connected by a load transmission device made of a laminated material of rubber and steel plates. Each of the reactor container and the outer building is disposed on a lower raft disposed on a rock by way of a vibration-proof device made of a laminated material of rubber and steel plates. Vibration-proof elements for providing vertical eigen frequency of the vibration-proof system comprising the reactor building and the vibration-proof device within a range of 3Hz to 5Hz are used. That is, the peak of designed acceleration for response spectrum in the horizontal direction of the reactor structural portions is shifted to side of shorter period from the main frequency region of the reactor structure. Alternatively, rigidity of the vibration-proof elements is decreased to shift the peak to the side of long period from the main frequency region. Designed seismic force can be greatly reduced both horizontally and vertically, to reduce the wall thickness of the structural members, improve the plant economy and to ensure the safety against earthquakes. (N.H.)

  6. Fuel assembly for FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki.

    1995-01-01

    Ordinary sodium bond-type fuel pins using nitride fuels, carbide fuels or metal fuels and pins incorporated with hydride moderators are loaded in a wrapper tube at a ratio of from 2 to 10% based on the total number of fuel pins. The hydride moderators are sealed in the hydride moderator incorporated pins at the position only for a range from the upper end to a reactor core upper position of substantially 1/4 of the height of the reactor core from the upper end of the reactor core as a center. Then, even upon occurrence of ULOF (loss of flow rate scram failure phenomenon), it gives characteristic of reducing the power only by a doppler coefficient and not causing boiling of coolant sodium but providing stable cooling to the reactor core. Therefore, a way of thinking on the assurance of passive safety is simplified to make a verification including on the reactor structure unnecessary. In an LMFBR type reactor using the fuel assembly, a critical experiment for confirming accuracy of nuclear design is sufficient for the item required for study and development, which provides a great economical effect. (N.H.)

  7. EPR, mu-Raman and Crystallographic properties of spinel type ZnCr{sub 2}O{sub 4}

    Energy Technology Data Exchange (ETDEWEB)

    Vargas-Hernandez, C; Almanza, O; Jurado, J F, E-mail: cvargash@unal.edu.c [Universidad Nacional de Colombia, Manizales-Colombia Laboratorio de Propiedades Opticas de Materiales-POM (Colombia)

    2009-05-01

    Structural, vibrational and electron paramagnetic resonance (EPR) analysis for compound ZnCr{sub 2}O{sub 4} are shown in this work. These types of materials are used in technological applications as humidity sensors and piezoelectric devices. The compound was obtained by mean of solid state reaction technique from binary precursors ZnO and Cr{sub 2}O{sub 3}. After three thermal treatments the sample structure was monitoring using X ray diffraction (XRD), the spinel cubic phase has been indexed within O{sup 7}{sub h}(Fd3m) spatial group. It is observed normal spinel phase. Micro-Raman analysis revealed bands for normal vibration modes of Zn and Cr atoms in tetrahedral and octahedral environments formed by oxygen atoms at approximately 400 and 900 cm{sup -1}, respectively. Bands around 941 cm{sup -1} are associated possibly to vacancies in the tetrahedral and octahedral sites due to interaction between Zn and Cr ions. EPR signal from 150 to 300 K isothermals indicates a transition between inverse spinel to normal spinel type in a central field around 3350 G. A signal at approximately 3400 G corresponding to the C'r{sup +3} in tetrahedral sites is observed near the central field.

  8. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  9. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  10. Standard test method for electrochemical reactivation (EPR) for detecting sensitization of AISI type 304 and 304L stainless steels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1994-01-01

    1.1 This test method covers a laboratory procedure for conducting an electrochemical reactivation (EPR) test on AISI Type 304 and 304L (UNS No. S30400 and S30403, respectively) stainless steels. This test method can provide a nondestructive means of quantifying the degree of sensitization in these steels (1, 2, 3). This test method has found wide acceptance in studies of the effects of sensitization on intergranular corrosion and intergranular stress corrosion cracking behavior (see Terminology G15). The EPR technique has been successfully used to evaluate other stainless steels and nickel base alloys (4), but the test conditions and evaluation criteria used were modified in each case from those cited in this test method. 1.2 The values stated in SI units are to be regarded as the standard. The inch-pound units given in parentheses are for information only. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this...

  11. EPR analysis of cyanide complexes of wild-type human neuroglobin and mutants in comparison to horse heart myoglobin.

    Science.gov (United States)

    Van Doorslaer, Sabine; Trandafir, Florin; Harmer, Jeffrey R; Moens, Luc; Dewilde, Sylvia

    2014-06-01

    Electron paramagnetic resonance (EPR) data reveal large differences between the ferric ((13)C-)cyanide complexes of wild-type human neuroglobin (NGB) and its H64Q and F28L point mutants and the cyanide complexes of mammalian myo- and haemoglobin. The point mutations, which involve residues comprising the distal haem pocket in NGB, induce smaller, but still significant changes, related to changes in the stabilization of the cyanide ligand. Furthermore, for the first time, the full (13)C hyperfine tensor of the cyanide carbon of cyanide-ligated horse heart myoglobin (hhMb) was determined using Davies ENDOR (electron nuclear double resonance). Disagreement of these experimental data with earlier predictions based on (13)C NMR data and a theoretical model reveal significant flaws in the model assumptions. The same ENDOR procedure allowed also partial determination of the corresponding (13)C hyperfine tensor of cyanide-ligated NGB and H64QNGB. These (13)C parameters differ significantly from those of cyanide-ligated hhMb and challenge our current theoretical understanding of how the haem environment influences the magnetic parameters obtained by EPR and NMR in cyanide-ligated haem proteins. Copyright © 2014 Elsevier B.V. All rights reserved.

  12. Some ideas on the EPR

    International Nuclear Information System (INIS)

    2003-01-01

    Facing the debate and controversial between partisans and opponents of the European Pressurized Reactor construction, the SFP energy Group aims to offer some reflexions. In this framework the following topics are discussed: the french nuclear park and its replacement, the energy costs, the nuclear reactors profitability, the generation IV reactors. The paper examines then the EPR technology and its cost to conclude on the advantage of an EPR construction, in the case of an energy policy based on the nuclear. This last point seems to be the real challenge of the problem. (A.L.B.)

  13. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  14. The E.P.R. (European Water Pressurized Reactor): a choice from the past that should close future?

    International Nuclear Information System (INIS)

    Marignac, Y.

    2003-11-01

    The present note discusses, by opening questions, different counter-arguments, of the French energy policy such the attempt to adapt the E.P.R., a technological project conceived in an old policy frame, to actual economical and environmental realities. (N.C.)

  15. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  16. EPR a strategic choice

    International Nuclear Information System (INIS)

    2003-01-01

    How can we answer to the increasing demand of electric power, resulting of the demographic evolution and needed to the economic development, without exhausting the fossil resources? The answers are function of the countries and imply an optimization of the production and the consumption. This document published by the Areva Group aims to show the advantages of the nuclear energy: economical and environmental advantages. A special chapter is devoted to the European Pressurized Reactor, EPR. (A.L.B.)

  17. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  18. Control rod for HTGR type reactor

    International Nuclear Information System (INIS)

    Mogi, Haruyoshi; Saito, Yuji; Fukamichi, Kenjiro.

    1990-01-01

    Upon dropping control rod elements into the reactor core, impact shocks are applied to wire ropes or spines to possibly deteriorate the integrity of the control rods. In view of the above in the present invention, shock absorbers such as springs or bellows are disposed between a wire rope and a spine in a HTGR type reactor control rod comprising a plurality of control rod elements connected axially by means of a spine that penetrates the central portion thereof, and is suspended at the upper end thereof by a wire rope. Impact shocks of about 5 kg are applied to the wire rope and the spine and, since they can be reduced by the shock absorbers, the control rod integrity can be maintained and the reactor safety can be improved. (T.M.)

  19. Method for controlling FBR type reactor

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

    1991-01-01

    The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

  20. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  1. Most advanced HTP fuel assembly design for EPR

    International Nuclear Information System (INIS)

    Francillon, Eric; Kiehlmann, Horst-Dieter

    2006-01-01

    End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

  2. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

    In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

  3. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  4. Communication channels analogous to one out of two oblivious transfers based on quantum uncertainty II Closing EPR-type loopholes

    CERN Document Server

    Shimizu, K

    2003-01-01

    In a recent proposal for a quantum cryptographic scheme analogous to one out of two oblivious transfers ÝPhys. Rev. A 66, 052316 (2002)¿, a sender, Bob, can encode two bits of information (X, Y) in a quantum carrier but a receiver, Alice, can decode only either X or Y dependent on her choice. Although Bob can discover her choice whenever he desires, she can detect this with a 50% probability. This paper clarifies the amount of information that Bob can expect to obtain from Alice without being detected by her by means of an Einstein-Podolsky-Rosen (EPR) type of attack. We can show that Bob inevitably fails to discover her choice with a 50% probability even though he can always completely escape her detection. (4 refs).

  5. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  6. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  7. Overview of LBB implementation for the EPR

    International Nuclear Information System (INIS)

    Cauquelin, C.

    1997-01-01

    This paper presents an overview of the use of leak-before-break (LBB) analysis for EPR reactors. EPR is an evolutionary Nuclear Island of the 4 loop x 1500 Mwe class currently in the design phase. Application of LBB to the main coolant lines and resulting design impacts are summarized. Background information on LBB analysis in France and Germany is also presented

  8. Overview of LBB implementation for the EPR

    Energy Technology Data Exchange (ETDEWEB)

    Cauquelin, C.

    1997-04-01

    This paper presents an overview of the use of leak-before-break (LBB) analysis for EPR reactors. EPR is an evolutionary Nuclear Island of the 4 loop x 1500 Mwe class currently in the design phase. Application of LBB to the main coolant lines and resulting design impacts are summarized. Background information on LBB analysis in France and Germany is also presented.

  9. EPR ohmic heating energy storage

    International Nuclear Information System (INIS)

    Heck, F.M.; Stillwagon, R.E.; King, E.I.

    1977-01-01

    The Ohmic Heating (OH) Systems for all the Experimental Power Reactor (EPR) designs to date have all used temporary energy storage to assist in providing the OH current charge required to build up the plasma current. The energies involved (0.8 x 10 9 J to 1.9 x 10 9 J) are so large as to make capacitor storage impractical. Two alternative approaches are homopolar dc generators and ac generators. Either of these can be designed for pulse duty and can be made to function in a manner similar to a capacitor in the OH circuit and are therefore potential temporary energy storage devices for OH systems for large tokamaks. This study compared total OH system costs using homopolar and ac generators to determine their relative merits. The total system costs were not significantly different for either type of machine. The added flexibility and the lower maintenance of the ac machine system make it the more attractive approach

  10. Control rod drives for FBR type reactor

    International Nuclear Information System (INIS)

    Ikakura, Hiroaki.

    1990-01-01

    The control rod drives for an FBR type reactor of the present invention eliminate obstacles deposited on attracting surfaces between an electromagnet and an armature which connect control rods to recover their retaining power. That is, a sealed chamber capable of controlling its inner pressure by an operation from the outside of a reactor is disposed in an extension pipe, and a nozzle connected to the sealed chamber and facing at the lower end thereof to the attracting surface is disposed. Liquid sodium sucked by evacuating the sealed chamber is jetted out from the nozzle by pressurizing the chamber to simultaneously eliminate obstacles deposited to the attracting surfaces of the electromagnet and the control rod. Alternatively, a nozzle protruding from and retracting to the lower surface of the electromagnet is disposed opposing to each of the attracting surfaces of the electromagnet and the control rod. Similar effect can also be obtained if gases are jetted out in this state. As a result, control rod drives of high reliability for a FBR type reactor can be obtained. (I.S.)

  11. Fueling method in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kawashima, Katsuyuki; Inoue, Kotaro.

    1985-01-01

    Purpose: To extend the burning cycle and decrease the number of fuel exchange batches without increasing the excess reactivity at the initial stage of burning cycles upon fuel loading to an LMFBR type reactor. Method: Each of the burning cycles is divided into a plurality of burning sections. Fuels are charged at the first burning section in each of the cycles such that driver fuel assemblies and blanket assemblies or those assemblies containing neutron absorbers such as boron are distributed in mixture in the reactor core region. At the final stage of the first burning section, the blanket assemblies or neutron absorber-containing assemblies present in mixture are partially or entirely replaced with driver fuel assemblies depending on the number of burning sections such that all of them are replaced with the driver fuel assemblies till the start of the final burning section of the abovementioned cycle. The object of this invention can thus be attained. (Horiuchi, T.)

  12. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  13. Steam generating system in LMFBR type reactors

    International Nuclear Information System (INIS)

    Kurosawa, Katsutoshi.

    1984-01-01

    Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

  14. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  15. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  16. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  17. EPR meeting international safety standards with margin

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Brauns, J.; Blombach, J.

    2005-01-01

    The EPR provides technology that offers a solution to the market's need for safe, economic power. The EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product capable of international licensing. As such, the EPR is a global product with commercial units currently being built in Finland at the Olkiluoto site, and planned for France, at the Flamanville site. Framatome ANP has recently proposed four EPR units to China in response to a request for vendor bids. In addition, Framatome ANP has announced their intent to pursue design certification with the United States Nuclear Regulatory Commission (NRC). This paper discusses how EPR's innovative safety philosophy ensures compliance with international safety standards for advanced light-water reactors (ALWRs). (author)

  18. EPR meeting international safety standards with margin

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Brauns, J.; Blombach, J.

    2005-01-01

    The EPR provides technology that offers a solution to the market's need for safe, economic power. The EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product capable of international licensing. As such, the EPR is a global product with commercial units currently being built in Finland at the Olkiluoto site, and planned for France, at the Flamanville site. Framatome ANP has recently proposed four EPR units to China in response to a request for vendor bids. In addition, Framatome ANP has announced their intent to pursue design certification in with the United States Nuclear Regulatory Commission (NRC). This paper discusses how EPR's innovative safety philosophy ensures compliance with international safety standards for advanced light-water reactors (ALWRs). (author)

  19. Technical description of other types of reactors

    International Nuclear Information System (INIS)

    Vollmer, H.

    1977-01-01

    The paper reviews the development of reactor systems other than LWR, i. e. gas cooled reactors, heavy water reactors and fast breeders. The specific features of these reactors are discussed. Technical details on plant design of the various systems will be given as well as the present status-of-the-art. (orig.) [de

  20. Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

    International Nuclear Information System (INIS)

    2015-01-01

    This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

  1. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  2. EPR: outlines of research and development

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    The articles give an overview of the research and development presently under way for the future pressurized water reactors to be constructed in Europe, and particularly the French and German EPR (European Pressurized Reactor). Such an analysis deals essentially with respect to the pressurized water reactors now in operation. (author)

  3. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

    Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems

  4. EPR design for maintenance

    International Nuclear Information System (INIS)

    Krugmann, U.

    1998-01-01

    Preventive maintenance is very important in achieving high plant availability. For the European Pressurized Reactor (EPR) preventive maintenance has been carefully addressed in the design stage. This is particularly necessary because of the traditionally different maintenance strategies employed in France and Germany. This paper emphasizes the following features introduced in the ERP design to minimize the duration of the refueling outage: (1) containment accessibility during power operation; (2) overall plant layout to facilitate inspections and maintenances within the containment; and (3) safety system design for enabling preventive maintenance during power operation. (author)

  5. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  6. Method of starting up PWR type reactor

    International Nuclear Information System (INIS)

    Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

    1988-01-01

    Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

  7. Self operation type reactor control device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1990-01-01

    A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

  8. Two-Photon Entanglement and EPR Experiments Using Type-2 Spontaneous Parametric Down Conversion

    Science.gov (United States)

    Sergienko, A. V.; Shih, Y. H.; Pittman, T. B.; Rubin, M. H.

    1996-01-01

    Simultaneous entanglement in spin and space-time of a two-photon quantum state generated in type-2 spontaneous parametric down-conversion is demonstrated by the observation of quantum interference with 98% visibility in a simple beam-splitter (Hanburry Brown-Twiss) anticorrelation experiment. The nonlocal cancellation of two-photon probability amplitudes as a result of this double entanglement allows us to demonstrate two different types of Bell's inequality violations in one experimental setup.

  9. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  10. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  11. Economical and engineering aspects of modular-type fast reactors

    International Nuclear Information System (INIS)

    Kirillov, E.V.; Demidova, L.S.

    1989-01-01

    Economical and engineering characteristics for SAFR and PRISM modular-type reactors are analyzed on the basis of foreign papers. Dependence of economical characteristics for SAFR modules on their output is shown. Cost of power generation for the NPPs with PRISM reactor, LWR reactor and for coal thermal power plant is presented

  12. Comparative study between two austenitic steels with the EPR (Electrochemical Potentiokinetic Reactivation) technique.; Estudio comparativo entre dos aceros austeniticos mediante la tecnica EPR (Reactivacion Electroquimica Potenciocinetica)

    Energy Technology Data Exchange (ETDEWEB)

    Guillen M, A N

    1997-09-01

    In the mid 19704s, the intergranular corrosion with stress corrosion cracking (IGSCC) have been identified as a greater problem in Boiling Water Reactors BWR in several places of the world. The Electrochemical Potentiokinetic Reactivation - Single Loop (EPR-SL) test and the Double Loop (EPR-DL) test, were developed as methods for measuring the Degree of Sensitization (DOS), show sensitised materials at subject to Intergranular Corrosion. In Mexico, the Laguna Verde4s reactor is BWR type and many of its principal components was built with AISI 304 stainless steels, while that in VVER reactors as well as Juragua4s reactor in Cuba is used 321 Stainless stell in its Russian equivalent designation 08Ch18N10T. In this work, were studied 304 and 08Ch18N10T stainless steels by means of EPR-SL, EPR-DL and ASTM A-262 techniques, they have been found a good correlation for 304 steel but not in 08Ch18N10T steel and was proposed one modification in the criterion by the evaluation on the sensitisation in this steels. Finally, both materials were welded with procedures used in the nuclear industry, by Slow Strain Rate Test (SSRT) to determine the Stress Corrosion Cracking SCC susceptibility, and subsequently the susceptibility to localized corrosion was studied by means of Cyclic Polarization test and the uniform corrosion rate in a solution with chlorides by the Tafel plot, Potentiodynamic Anodic Polarization Resistance. (Author).

  13. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

    2015-01-01

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  14. 75 FR 4882 - Meeting of the ACRS Subcommittee on EPR; Notice of Meeting

    Science.gov (United States)

    2010-01-29

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee On Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on EPR; Notice of Meeting The ACRS U.S. Evolutionary Power Reactor (EPR) Subcommittee will hold a... with Open Items concerning the U.S. EPR Design Certification (DCD) Application and the Calvert Cliffs...

  15. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  16. Non-electric applications of pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Adamov, E.O.; Cherkashov, Yu.M.; Romenkov, A.A.

    1997-01-01

    This paper recommends the use of pool-type light water reactors for thermal energy production. Safety and reliability of these reactors were already demonstrated to the public by the long-term operation of swimming pool research reactors. The paper presents the design experience of two projects: Apatity Underground Nuclear Heating Plant and Nuclear Sea-Water Desalination Plant. The simplicity of pool-type reactors, the ease of their manufacturing and maintenance make this type of a heat source attractive to the countries without a developed nuclear industry. (author). 6 figs, 1 tab

  17. A new type of EPR experiment using light quanta produced by nonlinear optical process

    International Nuclear Information System (INIS)

    Shih, Y.H.; Alley, C.O.

    1989-01-01

    A pair of correlated light quanta of 532 nm wavelength with the same linear polarization but divergent directions of propagation was produced by non-linear optical parametric down conversion in a crystal of deuterated potassium di-hydrogen phosphate from a 100 ps duration laser pulse of 266 nm wavelength. Each light quantum was converted to a circular polarization state or a linear polarization state (orthogonal) and was reflected by a turning mirror to superpose with the other at a beam splitter to produce a two-quanta superposition state. For coincident detection of the two light quanta at separated detectors, correlations of the Einstein-Podolsky-Rosen type for the polarizations have been observed as predicted by our analysis. In preliminary runs with limited data we have measured a violation of Bell's inequality by three standard deviations. We are planning to extend our experiments to include a truly random delayed choice between two analyser settings at each detector while maintaining a spacelike separation between the detections. (orig.)

  18. Simulation of a pool type research reactor

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  19. The market for HTGR type reactors

    International Nuclear Information System (INIS)

    Roehler, E.

    1986-01-01

    High-temperature-reactors with pebble-bed-reactor cores as a progressive reactor line, have been developed by BBC/HRB the Federal Republic of Germany over a period of 27 years and will soon be mature to be introduced to the market. They represent an important innovation in the field of reactor engineering. Due to its high degree of applicability on the power and heat market and its high flexibility regarding the site and fuel cycle the HTR is extremely suitable for providing energy to consumers, especially in countries using nuclear energy supply for the first time. (orig.) [de

  20. The E.P.R. (European Water Pressurized Reactor): a choice from the past that should close future?; L'EPR: un choix du passe qui fermerait l'avenir?

    Energy Technology Data Exchange (ETDEWEB)

    Marignac, Y

    2003-11-15

    The present note discusses, by opening questions, different counter-arguments, of the French energy policy such the attempt to adapt the E.P.R., a technological project conceived in an old policy frame, to actual economical and environmental realities. (N.C.)

  1. Simplified analysis of trasients in pool type liquid metal reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1987-01-01

    The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

  2. Steam generator of FBR type reactor

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1992-01-01

    Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

  3. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

  4. EPR and Bell Locality

    OpenAIRE

    Norsen, Travis

    2004-01-01

    A new formulation of the EPR argument is presented, one which uses John Bell's mathematically precise local causality condition in place of the looser locality assumption which was used in the original EPR paper and on which Niels Bohr seems to have based his objection to the EPR argument. The new formulation of EPR bears a striking resemblance to Bell's derivation of his famous inequalities. The relation between these two arguments -- in particular, the role of EPR as part one of Bell's two-...

  5. EPR dosimetry of irradiated human teeth

    International Nuclear Information System (INIS)

    Rodas Duran, J.E.; Panzeri, H.; Mascarenhas, S.

    1985-01-01

    The determination of the absorbed radiation dose in man may be made by Electron Paramagnetic Resonance (EPR) spectroscopy of dental enamel. We analysed the EPR signals for dental enamel submitted to gamma radiation in doses between 1 Gy and 25 Gy. We conclude that independent of the type of tooth analysed there exists a linear relation between the EPR signals and the absorbed doses. These studies were extended to enamel irradiated with gamma rays and with X rays in doses between 0.1 Gy and 0.6 Gy. The graph of the intensity of the EPR signals as a function of the dose has a slope of 0.22. This calibration may be used to calculate the absorbed dose for humans from a measurement of the EPR signal from small samples of enamel taken from any permanent tooth. Finally we comment on some EPR studies of effects of radiation of milk teeth. (author)

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

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

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

  7. Aging of reactor vessels in LWR type reactors

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-01-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs

  8. The working lifetime of nuclear power plants and new types of power reactors

    International Nuclear Information System (INIS)

    Bataille, Ch.; Birraux, C.

    2003-01-01

    The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

  9. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  10. EPR of alanine irradiated by neutrons

    International Nuclear Information System (INIS)

    Pivovarov, S.P.; Seredavina, T.A.; Zhdanov, S.V.; Mul'gin, S.I.; Zhakparov, R.K.

    2001-01-01

    In the work the first results of EPR studies of alanine, irradiated with diverse doses at neutron cyclotron generator different conditions and on the critical reactor stand are presented. A dose linearity dependence of EPR signal is observing, the methods of γ-background contribution separation are discussed. Obtain results is giving the basis to recommendation of alanine as an effective detector irradiation. However it is demanded the farther study on clarification of radiation sensitivity value dependence on the neutron energy spectrum form

  11. Comparative study between two austenitic steels with the EPR (Electrochemical Potentiokinetic Reactivation) technique

    International Nuclear Information System (INIS)

    Guillen M, A.N.

    1997-01-01

    In the mid 19704s, the intergranular corrosion with stress corrosion cracking (IGSCC) have been identified as a greater problem in Boiling Water Reactors BWR in several places of the world. The Electrochemical Potentiokinetic Reactivation - Single Loop (EPR-SL) test and the Double Loop (EPR-DL) test, were developed as methods for measuring the Degree of Sensitization (DOS), show sensitised materials at subject to Intergranular Corrosion. In Mexico, the Laguna Verde4s reactor is BWR type and many of its principal components was built with AISI 304 stainless steels, while that in VVER reactors as well as Juragua4s reactor in Cuba is used 321 Stainless stell in its Russian equivalent designation 08Ch18N10T. In this work, were studied 304 and 08Ch18N10T stainless steels by means of EPR-SL, EPR-DL and ASTM A-262 techniques, they have been found a good correlation for 304 steel but not in 08Ch18N10T steel and was proposed one modification in the criterion by the evaluation on the sensitisation in this steels. Finally, both materials were welded with procedures used in the nuclear industry, by Slow Strain Rate Test (SSRT) to determine the Stress Corrosion Cracking SCC susceptibility, and subsequently the susceptibility to localized corrosion was studied by means of Cyclic Polarization test and the uniform corrosion rate in a solution with chlorides by the Tafel plot, Potentiodynamic Anodic Polarization Resistance. (Author)

  12. Dynamic behaviour of a CAREM type reactor

    International Nuclear Information System (INIS)

    Abbate, P.; Doval, A.

    1990-01-01

    As complement to CAREM reactor design studies, behaviour analysis were made in a non-stationary regime, with the aim of developing plant systems and determining process variables variation ranges, characteristic of normal operating conditions, specifying alarm values for different variables, as well as for operating policies. Transient accidental scenes analysis were made, concluding that reactor characteristics provide security, maintaining the core integrity. (Author) [es

  13. Fuel exchange device for FBR type reactor

    International Nuclear Information System (INIS)

    Onuki, Koji.

    1993-01-01

    The device of the present invention can provide fresh fuels with a rotational angle aligned with the direction in the reactor core, so that the fresh fuels can be inserted being aligned with apertures of the reactor core even if a self orientation mechanism should fail to operate. That is, a rotational angle detection means (1) detects the rotational angle of fresh fuels before insertion to the reactor core. A fuel rotational angle control means (2) controls the rotational angle of the fresh fuels by comparing the detection result of the means (1) and the data for the insertion position of the reactor core. A fuel rotation means (3) compensates the rotational angel of the fresh fuels based on the control signal from the means (2). In this way, when the fresh fuels are inserted to the reactor core, the fresh fuels set at the same angle as that for the aperture of the reactor core. Accordingly, even if the self orientation mechanism should not operate, the fresh fuels can be inserted smoothly. As a result, it is possible to save loss time upon fuel exchange and mitigate operator's burden during operation. (I.S.)

  14. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  15. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  16. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  17. Risk reduction category (RRC-A) accident studies in the safety analysis report of the EPR trademark reactor

    International Nuclear Information System (INIS)

    Poehlmann, M.; Bleher, G.; Ismaier, A.; Knoll, A.; Levi, P.; Garcia, E. Vera; Schels, A.; Seitz, H.; Lima Campos, L.

    2013-01-01

    The Risk Reduction Category (RRC-A) is considered in the safety demonstration of nuclear reactors in addition to design basis operating conditions (Plant Condition Category, PCC), in order to analyze with a risk reduction approach any operating conditions with multiple failures. As extending the operating conditions of the plant 'beyond design basis', the Risk Reduction Category (RRC-A) is also denoted as Design Extension Condition (DEC-A). In the German licensing framework, the RRCA (or DEC-A) transients correspond to safety assessment level '4b' of the 'Sicherheitsanforderungen an Kernkraftwerke' (Safety Requirements for Nuclear Power Plants), Az. RS I 5 - 13303/01 of the German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety. These RRC-A (or DEC-A) operating conditions require specific design provisions (implemented by manual or automatic action), known as RRC-A measures, intended to render consequences of accumulated failures admissible. In contrast, RRC-B constitute severe accidents that lead to core melt. Identification of RRC-A operating conditions and corresponding RRC-A measures is based on the use of results of probabilistic safety assessments. After the Fukushima accident the RRC-A accidents like Station Black Out (SBO) or Loss of Ultimate Heat Sink (LUHS) are of particular interest in the safety assessment of nuclear new builds. In several chapters of the Safety Analysis Report it is demonstrated that the AREVA EPRTM design is resistant at RRC-A accident conditions. (orig.)

  18. EPR dosimetry in a mixed neutron and gamma radiation field.

    Science.gov (United States)

    Trompier, F; Fattibene, P; Tikunov, D; Bartolotta, A; Carosi, A; Doca, M C

    2004-01-01

    Suitability of Electron Paramagnetic Resonance (EPR) spectroscopy for criticality dosimetry was evaluated for tooth enamel, mannose and alanine pellets during the 'international intercomparison of criticality dosimetry techniques' at the SILENE reactor held in Valduc in June 2002, France. These three materials were irradiated in neutron and gamma-ray fields of various relative intensities and spectral distributions in order to evaluate their neutron sensitivity. The neutron response was found to be around 10% for tooth enamel, 45% for mannose and between 40 and 90% for alanine pellets according their type. According to the IAEA recommendations on the early estimate of criticality accident absorbed dose, analyzed results show the EPR potentiality and complementarity with regular criticality techniques.

  19. Confirmation of a second EPR to be built at Penly

    International Nuclear Information System (INIS)

    Anon.

    2009-01-01

    The French government has made public its decision to launch the construction of a second EPR in France. This EPR is planned to be built on the Penly site (northern region of France). EDF will manage this project but other investors like GDF-Suez are invited to participate. The construction works are planned to begin in 2012 for a commissioning in 2017. This reactor will be the fifth EPR being built in the world. (A.C.)

  20. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  1. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  2. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  3. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    Chichindaev, D.A.

    2001-01-01

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

  4. Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, 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, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  5. The EPR layout design

    International Nuclear Information System (INIS)

    Mast, U.; Le Carrer, P.Y.

    2001-01-01

    General: The European Pressurised Water Reactor (EPR) is a French - German development for the next generation of Pressurised Water Reactor. The new reactor design is based on the experiences of operation and design of nuclear power plants in both countries. The EPR fulfils enhanced safety standards, higher availability and a longer service life. Utilities aspects: For the Utilities one important requirement is the reduction of personnel exposure during maintenance and in-service inspection. The other significant requirement is of economic nature. The main points influencing costs, which have also impact on the layout, are: outage times, accessibility of the reactor building and the available maintenance and set down areas. The Utilities have also required to load the spent fuel assemblies into the shipping cask from the bottom of the fuel pool, because of the exclusion of the drop of the cask and in order to avoid contamination at the outer cask shell. Layout and safety aspects: All safety relevant Nuclear Island (NI) buildings are designed against design earthquake as well as explosion pressure wave. The protection against Airplane Crash (APC) is realised by civil and layout dispositions. The Reactor Building, the Safeguard Buildings division 2 and 3 and the Fuel Building are protected by concrete structures. The other safety relevant nuclear buildings are protected by geographical separation. Important safety requirements are the further reduction of the probability of severe accidents and the mitigation of such an accident on the plant area. For that, a spreading area for molten corium, a channel from the reactor pit to the spreading area and the In Containment Refuelling Water Storage Tank (IRWST) for flooding and initial cooling of the corium, were implemented in the design of the Reactor Building. Layout results: The following buildings are arranged on a common raft to protect them against design earthquake: Reactor Building (RB), Safeguard Buildings (SAB

  6. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  7. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  8. Core arrangement in BWR type reactors

    International Nuclear Information System (INIS)

    Asano, Masayuki.

    1981-01-01

    Purpose: To decrease the number of fuel assemblies whose locations are to be changed upon fuel exchange, as well as unify the power distribution in the core by arranging, in a chess board configuration, a plurality pattern of unit reactor lattices each containing fuel assemblies of different burnup degrees in orthogonal positions to each other. Constitution: A first pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 1 and fuel assemblies of burnup degree 3 at orthogonal positions to each other. A second pattern of unit reactor lattice is formed by disposing fuel assemblies of burnup degree 2 and fuel assemblies of burnup degree 1 at orthogonal positions to each other. The unit lattices each in such a dispositions are arranged in a chess board arrangement. Since, the fuel assemblies of the burnup degree 1 in the first pattern unit lattices proceed to the burnup degree 2 and the fuel assemblies of the burnup degree 2 in the second pattern unit lattices proceed to the burnup degree 3 up to the fuel exchange stage, fuel exchange and movement have only to be made, not for those fuel assemblies, but for another half of the fuel assemblies. (Kawakami, Y.)

  9. Primary cooling system for BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Eishi; Takahashi, Masanori; Aoki, Yasuko

    1993-01-01

    The present invention effectively uses information from a plurality of sensors in order to suppress corrosion circumstance of a nuclear reactor. That is, a predetermined general water quality factor at a predetermined position is determined as a standard index. A concentration of a water quality improver is controlled such that the index is within an aimed range. For this purpose, the entire sensor groups disposed in a primary coolant system of a nuclear reactor are divided into a plural systems of sensor groups each disposed on every different positions. Then, a predetermined sensor group (standard sensor group) is connected to a computing device and a data base so that it is always monitored for calculating and estimating the standard index. Only oxidative ingredient in water at the measuring point is noted, and a concentration distribution which agrees with an actually measured value of oxidative ingredients is extracted from data base and used as a correct concentration distribution. With such procedures, reactor water quality can be estimated accurately while compensating erroneous factors of individual sensors. Even when a new sensor is used, it is not necessary to greatly change control logic. (I.S.)

  10. Some elements of understanding about the cluster ejection accident in the EPR

    International Nuclear Information System (INIS)

    Vignon, Dominique

    2010-01-01

    The author answers to a publication made by an association (Sortir du Nucleaire) which is provided in appendix (some parts of this text are highlighted) and denounced risks associated with a cluster ejection accident in an EPR in relationship with steering modes which, according to this association, would be essentially related to an objective of economic profitability. The author first recalls some elements regarding the control and neutron stopping of pressurized water reactors. Then, after having outlined some specific aspects of the EPR design, he addresses the cluster ejection accident: safety approach and its application to this type of accident. He recalls the conclusions of studies of cluster ejection performed by EDF and AREVA, comments the consequences for the EPR power

  11. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  12. Self-operation type power control device for nuclear reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru.

    1993-01-01

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

  13. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, 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, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  14. The EPR: an unexplainable breach of the investment cost with respect to previous levels

    International Nuclear Information System (INIS)

    Dessus, Benjamin

    2012-01-01

    After having recalled some important figures published by the French Court of Audit in its report on the costs of the nuclear sector, the author comments the historic evolution of investment costs of the existing French nuclear reactors since the 1970's, and then the evolution of these costs for the different reactor types and generations. The author notably discusses the evolution of 'overnight' costs of the French nuclear reactors between 1978 and 2016, and outlines the huge increase noticed for the EPR

  15. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  16. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  17. Conceptual designs of power tokamak-type thermonuclear reactors

    International Nuclear Information System (INIS)

    Shejndlin, A.E.; Nedospasov, A.V.

    1978-01-01

    Physico-technical and ecological aspects of conceptual designing power tokamak-type reactors have been briefly considered. Only ''pure'' (''non-hybride'') reactors are discussed. Presented are main plasma-physical parameters, characteristics of blankets and magnetic systems of the following projects: PPPL; V-2; V-3; Culham-2, JAERI; TBEh-2500; TFTR. Two systems of the first wall protection have been considered: divertor one and by means of a layer of a cool turbulent plasma. Examined are the following problems: fuel loading, choice of the first wall material, blanket structure, magnetic system, environmental contamination. The comparison of relative hazards of fast neutron reactors and fusion reactors has shown that in respect of fusion reactors the biological hazard potential value is less by one-two orders

  18. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  19. Control rod drives for HTGR type reactor

    International Nuclear Information System (INIS)

    Nishiguchi, Isoharu; Katagiri, Shigeo.

    1991-01-01

    The device of the present invention has a feature of having stable braking characteristics upon scram operation of control rods. That is, control rod drives are moved upon and down by a dram which rotates the control rod suspended from to a wire rope, and the dram is disconnected from the driving mechanism by a crutch mechanism upon scram, to rapidly insert the control rod in the reactor by its own weight. An electric generator is used as a braking mechanism for controlling the scram speed of the control rod. A plurality of resistors disposed outside of the reactor coolants boundary are connected in parallel between input/output terminals of the electric generator. With such a constitution, braking characteristics are determined by the intensity of the permanent magnet, number of the coil windings and values of the resistors constituting the power generator. Accordingly, the braking characteristics are less changed relative to the working circumstantial conditions, the history of use and the state of mounting. As a result, stable braking characteristics can always be obtained. Further, braking characteristics can easily be controlled by varying the resistance value. (I.S.)

  20. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  1. Thermal shielding device in LMFBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Hiroshi.

    1985-01-01

    Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

  2. Self operation type reactor scram device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1992-01-01

    A control rod having neutron absorbers therein is held by a curie point electromagnet by way of a control rod extension shaft. The electromagnet is suspended from a vertically movable driving shaft in an upper guide tube. Then, a heater is disposed at the lower portion in the inner side of the upper guide tube. Upon a function confirmation test, the electromagnet is at first pulled up to the inside of the upper guide tube. Subsequently, the electromagnet is heated by the heater by a temperature higher than the curie point of the temperature sensing magnetic material. If the function is normal, armature connected to the control rod extension tube is separated. With such a constitution, the electromagnetic portion is isolated from a coolant main stream, thereby enabling to avoid the cooling effect by the stream of coolants. Accordingly, the operation test for confirming the integrity of the function of the curie point electromagnet can be conducted while placing the electromagnet in the reactor core as it is during actual reactor operation. (I.N.)

  3. BWR type reactor and its operating method

    International Nuclear Information System (INIS)

    Ootsuji, Niro.

    1983-01-01

    Purpose: To regulate the control rod extraction operation such that an assumed control rod drop accident, if should occur, may not lead to further serious accidents, as well as enable to improve the working life of the control rod. Method: A plurality of control rods disposed among a plurality of fuel assemblies constituting the reactor core for suppressing the reactor core reactivity are divided into two groups depending on the descending speed, and the number of rods with a faster descending speed is set to less than 1/4 of the total number of the control rods. Then, the control rods are arranged such that those rods of the faster descending speed may be set every one another in any of the vertical, lateral and orthogonal directions. Further, it is always judged as to the possibility of extracting the control rods with the faster descending speed by a fast control rod extraction judging circuit to issue a signal to a control rod extraction inhibition circuit, so that the extraction operation for the control rods with the faster descending speed is started after all of the control rods with the slow descending speed have been extracted. Accordingly, if a control rod dropping accident should occur, abrupt power change can be avoided to thereby minimize the development of the accident. (Horiuchi, T.)

  4. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

  5. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  6. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  7. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  8. Homogeneity and EPR metrics for assessment of regular grids used in CW EPR powder simulations.

    Science.gov (United States)

    Crăciun, Cora

    2014-08-01

    CW EPR powder spectra may be approximated numerically using a spherical grid and a Voronoi tessellation-based cubature. For a given spin system, the quality of simulated EPR spectra depends on the grid type, size, and orientation in the molecular frame. In previous work, the grids used in CW EPR powder simulations have been compared mainly from geometric perspective. However, some grids with similar homogeneity degree generate different quality simulated spectra. This paper evaluates the grids from EPR perspective, by defining two metrics depending on the spin system characteristics and the grid Voronoi tessellation. The first metric determines if the grid points are EPR-centred in their Voronoi cells, based on the resonance magnetic field variations inside these cells. The second metric verifies if the adjacent Voronoi cells of the tessellation are EPR-overlapping, by computing the common range of their resonance magnetic field intervals. Beside a series of well known regular grids, the paper investigates a modified ZCW grid and a Fibonacci spherical code, which are new in the context of EPR simulations. For the investigated grids, the EPR metrics bring more information than the homogeneity quantities and are better related to the grids' EPR behaviour, for different spin system symmetries. The metrics' efficiency and limits are finally verified for grids generated from the initial ones, by using the original or magnetic field-constraint variants of the Spherical Centroidal Voronoi Tessellation method. Copyright © 2014 Elsevier Inc. All rights reserved.

  9. Homogeneity and EPR metrics for assessment of regular grids used in CW EPR powder simulations

    Science.gov (United States)

    Crăciun, Cora

    2014-08-01

    CW EPR powder spectra may be approximated numerically using a spherical grid and a Voronoi tessellation-based cubature. For a given spin system, the quality of simulated EPR spectra depends on the grid type, size, and orientation in the molecular frame. In previous work, the grids used in CW EPR powder simulations have been compared mainly from geometric perspective. However, some grids with similar homogeneity degree generate different quality simulated spectra. This paper evaluates the grids from EPR perspective, by defining two metrics depending on the spin system characteristics and the grid Voronoi tessellation. The first metric determines if the grid points are EPR-centred in their Voronoi cells, based on the resonance magnetic field variations inside these cells. The second metric verifies if the adjacent Voronoi cells of the tessellation are EPR-overlapping, by computing the common range of their resonance magnetic field intervals. Beside a series of well known regular grids, the paper investigates a modified ZCW grid and a Fibonacci spherical code, which are new in the context of EPR simulations. For the investigated grids, the EPR metrics bring more information than the homogeneity quantities and are better related to the grids’ EPR behaviour, for different spin system symmetries. The metrics’ efficiency and limits are finally verified for grids generated from the initial ones, by using the original or magnetic field-constraint variants of the Spherical Centroidal Voronoi Tessellation method.

  10. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

    In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  11. Fuel assembly for BWR type reactor

    International Nuclear Information System (INIS)

    Kato, Shigeru.

    1993-01-01

    In the fuel assembly of the present invention, a means for mounting and securing short fuel rods is improved. Not only long fuel rods but also short fuel rods are disposed in channel of the fuel assembly to improve reactor safety. The short fuel rods are supported by a screw means only at the lower end plug. The present invention prevents the support for the short fuel rod from being unreliable due to the slack of the screw by the pressure of inflowing coolants. That is, coolant abutting portions such as protrusions or concave grooves are disposed at a portion in the channel box where coolants flowing from the lower tie plate, as an uprising stream, cause collision. With such a constitution, a component caused by the pressure of the flowing coolants is formed. The component acts as a rotational moment in the direction of screwing the male threads of the short fuel rod into the end plug screw hole. Accordingly, the screw is not slackened, and the short fuel rods are mounted and secured certainly. (I.S.)

  12. Double-walled tank type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishiguchi, Yohei.

    1993-01-01

    A secondary vessel containing a steam generator is disposed on a base slab, and a roof slab is disposed to the upper end opening of the base slab. A manometer sealing is formed between the upper end opening of the secondary vessel and the roof slab. A primary vessel is disposed in the second vessel for containing a reactor core therein. A communication pipeline system (equalizer) is disposed for communicating the cover gas space of the secondary vessel with the cover gas space of the primary vessel by way of the roof slab. The communication pipeline system comprises a breakable plate, a check valve which opens from the secondary system to the primary system, a closing valve and pipelines connecting them. Upon occurrence of a sodium-water reaction accident caused by rupture of heat transfer pipes of a steam generator in the secondary vessel, the breakable plate is broken to equalize the gas pressure by way of the communication pipelines. This can avoid external pressure buckling of the primary vessel. (I.N.)

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

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

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

  14. Fission rate measurements in fuel plate type assembly reactor cores

    International Nuclear Information System (INIS)

    Rogers, J.W.

    1988-01-01

    The methods, materials and equipment have been developed to allow extensive and precise measurement of fission rate distributions in water moderated, U-Al fuel plate assembly type reactor cores. Fission rate monitors are accurately positioned in the reactor core, the reactor is operated at a low power for a short time, the fission rate monitors are counted with detectors incorporating automated sample changers and the measurements are converted to fission rate distributions. These measured fission rate distributions have been successfully used as baseline information related to the operation of test and experimental reactors with respect to fission power and distribution, fuel loading and fission experiments for approximately twenty years at the Idaho National Engineering Laboratory (INEL). 7 refs., 8 figs

  15. Analysis of the consequences of the anomaly in the Flamanville EPR reactor pressure vessel head domes on their serviceability. Report to the Advisory Committee of Experts for Nuclear Pressure Equipment. Public version. Session of 26 and 27 June 2017

    International Nuclear Information System (INIS)

    CATTEAU, R.; HERVIOU, K.

    2017-06-01

    The Flamanville EPR reactor pressure vessel closure and bottom head domes were manufactured in 2006 and 2007 by forging in the Areva NP Creusot Forge plant. These components are subject to the technical qualification requirement of the ESPN order in reference because they present a risk of heterogeneity in their properties. For the purposes of this technical qualification, Areva NP measured bending rupture energy values lower than those mentioned in point 4 of appendix I to the ESPN order in reference [3], which led it in 2015 to propose an approach to ASN to demonstrate the adequate toughness of the material of these components, based on a program of testing on scale-one replica domes and mechanical assessments of the risk of fast fracture. This approach was examined by ASN and the French institute for radiation protection and nuclear safety (IRSN) and written up in the report in reference, was the subject of an opinion in reference of the Advisory Committee of experts for nuclear pressure equipment (GP ESPN), which met on 30 September 2015, and of ASN requests, more specifically concerning the in-service inspection provisions, in its letter in reference. Subject to these requests being taken into account, ASN considered that the demonstration approach is appropriate, provided that the phenomenon in question is identified and explained and that the data acquired through the test program are sufficient to characterise it. The first test results, in April 2016, led Areva NP to change its demonstration approach, notably the test program on scale-one replica domes, which gave rise to an information meeting with the GP ESPN on 24 June 2016, on the basis of the summary report drawn up by ASN and IRSN in reference. On the basis of the observations of the GP ESPN in reference, ASN informed Areva NP of additional requests in its letter in reference. The Areva NP test program was conducted for the most part in 2016. On 16 December 2016, Areva NP sent ASN a file in reference

  16. Long term review of research on light water reactor types

    International Nuclear Information System (INIS)

    Sumiya, Yutaka

    1982-01-01

    In Japan, 24 nuclear power plants of 17.18 million kWe capacity are in operation, and their rate of operation has shown the good result of more than 60% since 1980. One of the research on the development of light water reactors is the electric power common research, which was started in 1976, and 272 researches were carried out till 1982. It contributed to the counter-measures to stress corrosion cracking, thermal fatigue and the thinning of steam generator tubes, to the reduction of crud generation and the remote control and automation of inspection and maintenance, and to the verification of safety. The important items for the future are the cost down of nuclear power plant construction, the development of robots for nuclear power plants, the improvement of the ability to follow load variation, and the development of light water reactors of new types. It is necessary to diversify the types of reactors to avoid the effect of a serious trouble which may occur in one type of reactors. Tokyo Electric Power Co., Inc., thinks that the Japanese type PWRs having the technical features of KWU type PWRs are desirable for the future development. The compatibility with the condition of installation permission in Japan, the required design change and the economy of the standard design PWRs of KWU (1.3 million kW) have been studied since October, 1981, by KWU and three Japanese manufacturers. (Kako, I.)

  17. Should the EPR be financed by industrialists?

    International Nuclear Information System (INIS)

    Jemain, A.

    2005-01-01

    As a reciprocal arrangement, the industrialists should be invited to contribute to the financing of the new generation reactor EPR (European pressurized reactor) which will be built in Flamanville (Manche, France). In exchange to their financial participation, the industrialists will receive the contractual warranty of stable electricity prices during 10 to 15 years periods. This short paper presents the opposite opinion of two representative of French industries concerning this proposal. Short paper. (J.S.)

  18. The EPR operators are trained on simulator

    International Nuclear Information System (INIS)

    Maincent, G.

    2009-01-01

    Three years before the EPR reactor of Flamanville (Normandie, France) is generating its very first kilowatt hours, Electricite de France has started to train its teams on a simulator which reproduces the man-machine interface of the future nuclear power plant. The simulator used is an evolutive tool specific to the Flamanville reactor and capable to test about 20 different accidental situations. (J.S.)

  19. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  20. Evaluation of plate type fuel options for small power reactors

    International Nuclear Information System (INIS)

    Andrzejewski, Claudio de Sa

    2005-01-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  1. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  2. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  3. Mechanical design and analysis for a EPR first wall/blanket/shield system

    International Nuclear Information System (INIS)

    Stevens, H.C.; Misra, B.; Youngdahl, C.K.

    1978-01-01

    Continuing studies are in progress at ANL to expand upon the design of a first wall/blanket/shield FW/B/S system and power conversion for a tokamak type Experimental Power Reactor (EPR). The FW/B/S system has evolved from an earlier design for a low beta, circular cross section plasma (major radius = 6 m) to one for a higher beta elongated plasma with a 4.7 m major radius. Basic mechanical design and layout features of the old and new EPR designs showing some of the more important design developments are given. These developments are aimed at simplifying the design, reducing the costs and in addition, improving the plant thermal efficiency and overall maintainability. In the area of the reactor blanket, significant thermal hydraulic and stress analysis have been performed to substantiate the integrity of the chosen concept. This paper deals with the discussion of these improved features

  4. Mast-sipping in EPR trademark plants

    International Nuclear Information System (INIS)

    Langenberger, Jan; Schienbein, Marcel; Geier, Roland

    2010-01-01

    For more than 20 years, AREVA applies and develops different sipping techniques to identify fuel assemblies with leaking fuel rods. For the EPR trademark reactors a Mast Sipping System with newest developments will be implemented considering radiation protection and latest standards requirements. The innovative EPR trademark Sipping System differs from previous systems in many ways. One of the main innovations is that all the necessary processes of the Sipping system have been fully digitized. Second, several ALARA design modifications have been implemented to meet the current radiation protection requirements. An additional implementable multilingual assistance program facilitates the handling of the system and helps to prevent incorrect operation. (orig.)

  5. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  6. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  7. New generation nuclear power units of PWR type integral reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

    1997-01-01

    Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

  8. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

    Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

  9. Medical application of EPR

    International Nuclear Information System (INIS)

    Eichhoff, Uwe; Hoefer, Peter

    2015-01-01

    Selected applications of continuous-wave EPR in medicine are reviewed. This includes detection of reactive oxygen and nitrogen species, pH measurements and oxymetry. Applications of EPR imaging are demonstrated on selected examples and future developments to faster imaging methods are discussed

  10. Reactor design and safety approach for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Davies, S.M.; Yamaki, Hideo; Goodman, L.

    1984-06-01

    A tank type plant has been designed that offers compactness, high reliability under seismic and thermal transients, and a safety design approach that provides a balance between public safety and plant availability. This report provides a description of the design philosophy and safety features of the reactor

  11. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  12. Description of the magnox type of gas cooled reactor (MAGNOX)

    International Nuclear Information System (INIS)

    Jensen, S.E.; Nonboel, E.

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO 2 ) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  13. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1983-01-01

    This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

  14. Artificial intelligence applied to fuel management in BWR type reactors

    International Nuclear Information System (INIS)

    Ortiz S, J.J.

    1998-01-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  15. Perspective channel-type reactor with enhanced safety

    International Nuclear Information System (INIS)

    Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

    1994-01-01

    Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

  16. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Sang-Guk

    2014-01-01

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

  17. Sensitivity analysis of reflector types and impurities in 10 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Khamis, I.

    2007-01-01

    The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

  18. Sensitivity analysis of reflector types and impurities in a 10 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Khamis, I.

    2008-01-01

    The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

  19. The long way to the EPR

    International Nuclear Information System (INIS)

    Berke, C.

    1997-01-01

    At the joint conference organized by the German Kerntechnische Gesellschaft (KTG) e.V. and the French Societe Francaise d'Energie Nucleaire (SFEN) in Cologne, the announcement was made on October 20, 1997 that Jean Claude Leny will be appointed Honorary Member of KTG. For reasons of ill health the new honorary member was prevented from attending the conference. In his place, Claus Berke sketched the history of the European Pressurized Water Reactor (EPR) and presented an overview of nuclear development in France and in Germany. Developing one standard reactor model in the primary objective of German-French cooperation in reactor development. The EPR is to replace the present nuclear generating units in France and Germany after the end of their planned service life, and is to be marketed as a standard export line. (orig.) [de

  20. Proven commercial reactor types: an introduction to their principal advantages and disadvantages

    International Nuclear Information System (INIS)

    Alesso, H.P.

    1981-01-01

    This study deals with the principal advantages and disadvantages of the five types of proven commercial reactors. A description of each class of commercial reactor (light water, gas-cooled, and heavy water) and their proven reactors is followed by a comparison of reactor types on the basis of technical merit, economics of operation, availability of technology, and associated political issues. (author)

  1. Overall aspects of control of ISIS-type nuclear reactor

    International Nuclear Information System (INIS)

    Amato, S.; Santinelli, A.

    1996-01-01

    The paper describes the main aspects related to the definition of main controls required to operate an ISIS-type nuclear power reactors. ISIS is a PWR-type intrinsically safe nuclear reactor designed by ANSALDO, based on density lock concept; it presents, between the other safety functions, self-depressurization and core cooling capability for unlimited time. Due to its specific characteristics, the ISIS reactor required to development of new control philosophy (if compared with actual nuclear power reactor) with the implementation of new control functions, for instance the density locks hot/cold interface locations control. This paper describes the main control functions implemented, their rationale, as well as the dynamic simulation performed to verify the adequacy of controls definitions. The dynamic simulations here described refers to a step-wise power ramp of 100-90-100 (% of nominal power) and to a power ramp of 100-50-100 with a slope of 5%/min; the results obtained have shown the ISIS capability to perform such operational transients, despite its innovative design was mainly focused on intrinsically safe behaviour. (author)

  2. One-way EPR steering and genuine multipartite EPR steering

    Science.gov (United States)

    He, Qiongyi; Reid, Margaret D.

    2012-11-01

    We propose criteria and experimental strategies to realise the Einstein-Podolsky-Rosen (EPR) steering nonlocality. One-way steering can be obtained where there is asymmetry of thermal noise on each system. We also present EPR steering inequalities that act as signatures and suggest how to optimise EPR correlations in specific schemes so that the genuine multipartite EPR steering nonlocality (EPR paradox) can also possibly be realised. The results presented here also apply to the spatially separated macroscopic atomic ensembles.

  3. Costing the EPR Project Using the Real Options Method

    International Nuclear Information System (INIS)

    Epaulard, Anne; Gallon, Stephane

    2001-01-01

    Real options theory makes it possible to cost investments which offer flexibility but whose returns are uncertain, such as the construction in 2000 of an EPR prototype; this prototype will enable the European pressurised-water reactor (EPR) to be used to renew EDF's nuclear power stations in 2020 (flexibility) but its economic worth will then depend on the cost of the competing gas-fired power plants (uncertain return). Options theory shows that investing in EPR technology in 2000 provides sufficient flexibility in 2020 to be considered cost-effective, even though use of EPRs is unlikely by that date. The investment made in 2000 to develop EPR technology therefore actually plays the part of an option or, in other words, insurance (against the risk of high gas prices)

  4. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  5. Raw materials problems in connection with fast breeder type reactors

    International Nuclear Information System (INIS)

    Hirsch, H.; Kreusch, J.

    1981-01-01

    The power supply by the FBR type reactors which depends upon the availability of essential raw materials such as Cr and Mo for structural and special steels is supposed to be less ensured than supply by fossil-fueled power plants. This contribution tries to verify this statement by means of estimates of the annual Cr and Mo demand, of the resources, production and consumption as well as by a study of the possibilities of recycling and substituting Cr and Mo. The only realistic alternative to the fast breeder type reactor is supposed to be a soft path of development according to the principle of decentralization, utilization of renewable energy sources regard to environmental protection, and use of less sophisticated technology. (DG) [de

  6. MDEP Common Position No EPR-01 - Common positions on the EPR instrumentation and controls design

    International Nuclear Information System (INIS)

    2010-01-01

    The purpose of the EPR Working Group (EPRWG) of the Multinational Design Evaluation Program (MDEP) is to identify common positions among the regulators reviewing the EPR Instrumentation and Controls (I and C) Systems in order to: 1. Promote understanding of each country 's regulatory decisions and basis for the decisions, 2. Enhance communication among the members and with external stakeholders, 3. Identify areas where harmonization and convergence of regulations, standards, and guidance can be achieved or improved, and 4. Supports standardization of new reactor designs. Since January 2008, the EPR I and C Technical Expert Subgroup (TESG) members met five times to exchange information regarding their country 's review of the EPR I and C design. The EPR I and C TESG consists of regulators from China, Canada, Finland, France, the United Kingdom, and the United States. The information exchange includes presentation of each country 's review status and technical issues, sharing of guidance documents, and sharing of regulatory decision documents. The TESG focused on the following four core areas of the EPR I and C design: 1. I and C System Independence (particularly for data communications), 2. Level of Defense and Diversity (back-up systems), 3. Qualification/quality of digital platforms, 4. Categorization/classification of systems and functions. As meetings were conducted, some areas were emphasized more depending on the significance of the issues for each country. During the TESG interactions, it became apparent that there were aspects of the EPR design where the countries had common agreement. On November 2, 2009, three of the subgroup countries, France, Finland and the United Kingdom, issued a joint regulatory position on the EPR I and C design as result of the 'Groupe Permanent' meeting in France. This statement of common positions expands upon that joint regulatory position

  7. Temperature fluctuation reducing device for FBR type reactor

    International Nuclear Information System (INIS)

    Ootsuka, Fumio; Shiratori, Fumihiro.

    1991-01-01

    In existent FBR type reactors, since temperature fluctuation in the reactor upper portion has been inevitable, thermal fatigue may be caused possibly in reactor core upper mechanisms. Then, a valve is disposed to a control rod lower guide tube contained in a reactor container for automatically controlling the amount of passing coolants in accordance with the temperature of the passing coolants, to mix and control coolants passing through a fuel assembly in adjacent with the guide tube and coolants passing through the guide tube. Further, a rectification cylinder is disposed, in which a portion of coolants passing through the fuel assembly is caused to flow. An orifice is disposed to the cylinder with an exit being disposed to the upstream thereof such that the coolants not flown into the rectification cylinder and the coolants passing through the guide tube are mixed to moderate the temperature fluctuation. That is, a portion of the coolants flown into the rectification cylinder can not pass through the orifice, but flow backwardly to the upstream and is discharged out of the rectification cylinder from the coolants exit and mixed sufficiently with coolants passing through the guide tube. In this way, temperature fluctuation can be moderated. (N.H.)

  8. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  9. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  11. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Fischer, M.; Bittermann, D.

    2005-01-01

    The EPR, an evolutionary pressurized water reactor (PWR), is a 4300-4500 MWth that incorporates proven technology within an optimized configuration to enhance safety. EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product. Commercial EPR units are currently being built in Finland at the Olkiluoto site, and planned for France at the Flamanville site. In recent months, Framatome ANP announced their intention to market the EPR units to China in response to a request for vendor bids as well as their intent to pursue design certification in the United States under 10CFR52. The EPR safety philosophy is based on a deterministic consideration of defense-in-depth complemented by probabilistic analyses. Not only is the EPR designed to prevent and mitigate design basis accidents (DBAs), it employs an extra level of safety associated with severe accident response. Therefore, as a design objective, features are included to ensure that radiological consequences are limited such that the need for stringent counter measures, such as evacuation and relocation of the nearby population, can be reasonably excluded. This paper discusses some of the innovative features of the EPR to address severe accident challenges. (author)

  12. Nuclear choice. Towards the EPR with mincing steps

    International Nuclear Information System (INIS)

    Jemain, A.

    2003-01-01

    The French government is on the way to give its green light for the construction of the EPR (European Pressurized water Reactor), the 1600 MWe demonstrator reactor of third generation of Framatome ANP (Areva). Fifty seven months will be necessary for its construction which, with the public inquiry, would lead to a first commissioning by 2010. This reactor will progressively renew the present day park between 2015-2020. The EPR project will cost 3 billions of euros but the financing remains to be found. Short paper. (J.S.)

  13. EPR trademark project delivery. The value of experience

    International Nuclear Information System (INIS)

    Leverenz, Ruediger

    2013-01-01

    We are building the EPR trademark reactor fleet. Together. With four EPR trademark projects under construction in the world, AREVA has unrivalled experience in the delivery of large-scale nuclear projects, including more than a thousand lessons learned captured from Olkiluoto 3 and Flamanville 3 projects. This book of knowledge as well as the return of experience of AREVA's and EDF's teams are now being fully leveraged on ongoing projects, especially on Flamanville 3 and Taishan, and will be incorporated in all future EPR TM projects.

  14. The EPR paradox revisited

    International Nuclear Information System (INIS)

    Cantrell, C.D.; Scully, M.O.

    1978-01-01

    Einstein, Podolsky and Rosen (EPR) argued in 1935 that quantum mechanics fails to give an adequate description of physical reality, and also cannot give a consistent wave-function description of certain phenomena. The authors show that a calculation based upon the reduced density matrix removes the formal inconsistency pointed out by EPR. The spirit of the present paper is that of a pedagogical review. (Auth.)

  15. EPR paradox revisited

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, C.D.; Scully, M.O.

    1978-07-01

    Einstein, Podolsky, and Rosen (EPR) argued in 1935 that quantum mechanics fails to give an adequate description of physical reality, and also cannot give a consistent wave-function description of certain phenomena. It is shown that a calculation based upon the reduced density matrix removes the formal inconsistency pointed out by EPR. The spirit of the present paper is that of a pedagogical review.

  16. Influence of glycemic control on some real-time biomarkers of free radical formation in type 2 diabetic patients: An EPR study.

    Science.gov (United States)

    Gadjeva, Veselina Georgieva; Goycheva, Petia; Nikolova, Galina; Zheleva, Antoaneta

    2017-11-01

    The pathology of diabetes is associated with several mechanisms, one of which is oxidative stress (OS). The relationship between OS and diabetic complications has been extensively investigated. OS has been suggested to be involved in the genesis of both macroand microangiopathy. In contrast, the relationship between OS and insulin action is a neglected research area. The aim of this study is to elucidate the effect of glycemic control in type 2 diabetic patients by following the serum levels of some real-time oxidative stress biomarkers. The study group consisted of 53 type 2 diabetic patients (31 with poor glycemic control and 22 with good glycemic control) and 24 healthy control subjects. The oxidative stress biomarkers (ROS, Asc• and •NO) were measured by using electron paramagnetic resonance spectroscopy (EPR) methods and compared with clinical parameters. The statistically significantly higher levels of ROS products and •NO in type 2 diabetic patients in both groups compared to controls mean that the oxidation processes take place at the time the survey is performed. Free radical overproduction persists after the normalization of the glucose levels, and oxidative stress may be involved in the "metabolic memory" effect. This is confirmed by the positive correlation between ROS levels/•NO and average blood glucose levels, triglycerides, and total cholesterol. Furthermore, the low level of the ascorbate radical in both diabetes groups compared to controls confirmed an increase in oxidation processes. Higher levels of real-time biomarkers show that intensive insulin treatment does not lead to the expected decrease in oxidative processes involving ROS and •NO, probably due to "metabolic memory".

  17. Emergency core cooling system for LMFBR type reactors

    International Nuclear Information System (INIS)

    Tamano, Toyomi; Fukutomi, Shigeki.

    1980-01-01

    Purpose: To enable elimination of decay heat in an LMFBR type reactor by securing natural cycling force in any state and securing reactor core cooling capacity even when both an external power supply and an emergency power supply are failed in emergency case. Method: Heat insulating material portion for surrounding a descent tube of a steam drum provided at high position for obtaining necessary flow rate for flowing resistance is removed from heat transmitting surface of a recycling type steam generator to provide a heat sink. That is, when both an external power supply and an emergency power supply are failed in emergency, the heat insulator at part of a steam generator recycling loop is removed to produce natural cycling force between it and the heat transmitting portion of the steam generator as a heat source for the heat sink so as to secure the flow rate of the recycling loop. When the power supply is failed in emergency, the heat removing capacity of the steam generator is secured so as to remove the decay heat produced in the reactor core. (Yoshihara, H.)

  18. Accident transient processes at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    1982-01-01

    Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

  19. Reactor building seismic analysis of a PWR type - NPP

    International Nuclear Information System (INIS)

    Kakubo, Masao

    1983-01-01

    Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

  20. Analysis of the procedure proposed by AREVA to prove adequate toughness of the domes of the Flamanville 3 EPR reactor pressure vessel (RPV) lower head and closure head. Session of 30 September 2015. Public version

    International Nuclear Information System (INIS)

    Catteau, R.; Cadet-Mercier, S.

    2015-01-01

    AREVA has asked ASN to evaluate the conformity of the reactor pressure vessel (RPV) for the Flamanville 3 EPR in application of the order reference [6]. The domes of the Flamanville 3 RPV closure head and lower head were manufactured in 2006 and 2007. AREVA identified that these components displayed a risk of heterogeneity of their characteristics and therefore carried out a technical qualification. At the end of 2014, AREVA informed ASN of lower-than-expected results of impact tests conducted as part of this technical qualification on test specimens taken from a dome representative of those intended for Flamanville 3. The values measured on two series of three test specimens give a mean value of 52 joules which does not attain the quality standard expected by AREVA. This mean value is also lower than the bending rupture energy value of 60 joules mentioned in point 4 of appendix 1 of the order reference [6], with which compliance would have been sufficient to prove the toughness of the material. AREVA carried out investigations to determine the origin of these noncompliant values. The carbon concentration measurements taken at the surface of the representative dome by portable spectrometry revealed the presence of a zone of major positive segregation (high concentration of carbon) over a diameter of about one meter. Furthermore, the examinations show that the segregation extends to a depth exceeding a quarter of the thickness of the dome. AREVA explains the non-compliance with the bending rupture energy criterion by the presence of this major positive segregation which came from the ingot used for the forging and was not completely eliminated by the cropping operations. To deal with this deviation, AREVA plans proving that the material is sufficiently tough by conducting new tests on a material that is representative of the lower and upper domes of the Flamanville EPR reactor. The body of the Flamanville 3 RPV, of which the lower dome is a part, has already

  1. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  2. Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

    International Nuclear Information System (INIS)

    Permana, Sidik

    2009-01-01

    Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

  3. The Finnish EPR dependent on a bridge

    International Nuclear Information System (INIS)

    Dupin, L.

    2010-01-01

    A new conflict has emerged between the main contractor Areva and its client TVO, the Finnish energy group. The qualification of a handling bridge has delayed the OL3 project of third EPR reactor at Olkiluoto, Finland. The disagreement concerns the testing procedure and may induce a two year delay in the completion of the project which is now planned for 2012 instead of 2010 initially. (J.S.)

  4. Case Studies in e-RPL and e-PR

    Science.gov (United States)

    Cameron, Roslyn; Miller, Allison

    2014-01-01

    The use of ePortfolios for recognition of prior learning (e-RPL) and for professional recognition (e-PR) is slowly gaining in popularity in the VET sector however their use is sporadic across educational sectors, disciplines, educational institutions and professions. Added to this is an array of purposes and types of e-RPL and e-PR models and…

  5. The utility of different reactor types for the research

    International Nuclear Information System (INIS)

    Stiennon, G.

    1983-01-01

    The report presents a general view of the use of the different belgian research reactor i.e. venus reactor, BR-1 reactor, BR-2 reactor and BR-3 reactor. Particular attention is given to the programmes which is in the interest of international collaboration. In order to reach an efficient utilization of such reactors they require a specialized personnel groups to deal with the irradiation devices and radioactive materials and post irradiation examinations, creating a complete material testing station. (A.J.)

  6. Method of freezing type dismantling for wasted reactors

    International Nuclear Information System (INIS)

    Tatsumi, Toshiyuki.

    1985-01-01

    Purpose: To enable to operate a cutting device in the air by placing a working table on ice while utilizing the ice as radiation shielding materials thereby prevent the diffusion of air contaminations. Method: Upon dismantling a BWR type reactor, ice is packed into a reactor container and a pressure vessel and frozen state is maintained by cooling coils disposed to the outer circumference of the pressure vessel. Then, an airtight hood is covered over the pressure vessel and a working table is rotatably disposed therein. Upon working, when the upper layer ice is melted by a heat pump and discharged, the airtight hood is lowered to a predetermined level. After freezing the melted portion again at the lowered level, cutting work is conducted by an operator in the hood. The cut pieces are conveyed after hoisting the airtight hood by a crane. The pressure vessel is dismantled by repeating the foregoing procedures. In this way, cut pieces can be recovered without falling them to the reactor bottom as in the conventional work in water. In addition, since the procedures are conducted while covering the airtight hood, diffusion of air contaminations can be prevented. (Kamimura, M.)

  7. Saturated steam turbines for power reactors of WWER-type

    International Nuclear Information System (INIS)

    Czwiertnia, K.

    1978-01-01

    The publication deals with design problems of large turbines for saturated steam and with problem of output limitations of single shaft normal speed units. The possibility of unification of conventional and nuclear turbines, which creates the economic basis for production of both types of turbines by one manufacturer based on standarized elements and assemblies is underlined. As separate problems the distribution of nuclear district heating power systems are considered. The choice of heat diagram for district heating saturated steam turbines, the advantages of different diagrams and evaluaton for further development are presented. On this basis a program of unified turbines both condensing and district heating type suitable for Soviet reactors of WWER-440 and WWER-1000 type for planned development of nuclear power in Poland is proposed. (author)

  8. Novel homo- and heterobinuclear ball-type phthalocyanines: synthesis and electrochemical, electrical, EPR and MCD spectral properties.

    Science.gov (United States)

    Odabaş, Zafer; Dumludağ, Fatih; Ozkaya, Ali Riza; Yamauchi, Seigo; Kobayashi, Nagao; Bekaroğlu, Ozer

    2010-09-21

    The mononuclear Fe(II) phthalocyanine 2 and ball-type homobinuclear Fe(II)-Fe(II) and Cu(II)-Cu(II) phthalocyanines, 3 and 4 respectively, were synthesized from the corresponding 4,4'-[1,1'-methylenebis-(naphthalene-2,1-diyl)]bis(oxy)diphthalonitrile 1, and then ball-type heterobinuclear Fe(II)-Cu(II) phthalocyanine 5 was synthesized from 2. The novel compounds 4 and 5 have been characterized by elemental analysis, UV/vis, IR and MALDI-TOF mass spectroscopies. Electron paramagnetic resonance and magnetic circular dichroism measurements of 3, 4 and 5 were also examined. The voltammetric measurements of the complexes showed the formation of various electrochemically stable ligand- and metal-based mixed-valence species, due to the intramolecular interactions between the two MPc units, especially in ball-type binuclear iron(II) phthalocyanine. Impedance spectroscopy and d.c. conductivity measurements of 4 and 5 were performed as a function of temperature (295-523 K) and frequency (40-10(5) Hz). While room temperature impedance spectra consist of a curved line, a transformation into a full semicircle with increasing temperature was observed for both compounds.

  9. Thermalhydraulics of flowing particle-bed-type fusion reactor blankets

    International Nuclear Information System (INIS)

    Nietert, R.E.; Abdelk-Khalik, S.I.

    1982-01-01

    An experimental investigation has been conducted to determine the heat transfer characteristics of gravity-flowing particle beds using a special heat transfer loop. Glass microspheres were allowed to flow by gravity at controlled rates through an electrically heated stainless steel tubular test section. Values of the local and average convective heat transfer coefficient as a function of the average bed velocity, particle size and heat flux were determined. Such information is necessary for the design of gravity-flowing particle-bed type fusion reactor-blankets and associated tritium recovery systems. (orig.)

  10. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  11. WWER type reactors used as multipurpose nuclear power sources

    International Nuclear Information System (INIS)

    Fiala, J.; Mulak, J.

    1976-01-01

    Safety aspects are assessed of the siting of nuclear power installations in the vicinity of large housing estates and in areas with a high population density, mainly the aspect of the liquidation of the consequences of the maximum credible accident, i.e., the transversal rupture of the primary coolant circuit. The application of WWER type reactors as multipurpose nuclear power sources in Czechoslovakia is justified. It is shown that such a multipurpose nuclear power source differs from a purely condensation nuclear power plant mainly in the design of the secondary stage. The possibilities of such projects are indicated with a view to power and heat operation. (F.M.)

  12. Life time of nuclear power plants and new types of reactors

    International Nuclear Information System (INIS)

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  13. Biomolecular EPR spectroscopy

    CERN Document Server

    Hagen, Wilfred Raymond

    2008-01-01

    Comprehensive, Up-to-Date Coverage of Spectroscopy Theory and its Applications to Biological SystemsAlthough a multitude of books have been published about spectroscopy, most of them only occasionally refer to biological systems and the specific problems of biomolecular EPR (bioEPR). Biomolecular EPR Spectroscopy provides a practical introduction to bioEPR and demonstrates how this remarkable tool allows researchers to delve into the structural, functional, and analytical analysis of paramagnetic molecules found in the biochemistry of all species on the planet. A Must-Have Reference in an Intrinsically Multidisciplinary FieldThis authoritative reference seamlessly covers all important bioEPR applications, including low-spin and high-spin metalloproteins, spin traps and spin lables, interaction between active sites, and redox systems. It is loaded with practical tricks as well as do's and don'ts that are based on the author's 30 years of experience in the field. The book also comes with an unprecedented set of...

  14. The water desalination complex based on ABV-type reactor plant

    International Nuclear Information System (INIS)

    Panov, Yu.K.; Fadeev, Yu.P.; Vorobiev, V.M.; Baranaev, Yu.D.

    1997-01-01

    A floating nuclear desalination complex with two barges, one for ABV type reactor plant, with twin reactor 2 x 6 MW(e), and one for reverse osmosis desalination plant, was described. The principal specifications of the ABV type reactor plant and desalination barge were given. The ABV type reactor has a traditional two-circuit layout using an integral type reactor vessel with all mode natural convection of primary coolant. The desalted water cost was estimated to be around US $0.86 per cubic meter. R and D work has been performed and preparations for commercial production are under way. (author)

  15. Fuel loading method to exchangeable reactor core of BWR type reactor and its core

    International Nuclear Information System (INIS)

    Koguchi, Kazushige.

    1995-01-01

    In a fuel loading method for an exchangeable reactor core of a BWR type reactor, at least two kinds of fresh fuel assemblies having different reactivities between axial upper and lower portions are preliminarily prepared, and upon taking out fuel assemblies of advanced combustion and loading the fresh fuel assemblies dispersingly, they are disposed so as to attain a predetermined axial power distribution in the reactor. At least two kinds of fresh fuel assemblies have a content of burnable poisons different between the axial upper portion and lower portions. In addition, reactivity characteristics are made different at a region higher than the central boundary and a region lower than the central boundary which is set within a range of about 6/24 to 16/24 from the lower portion of the fuel effective length. There can be attained axial power distribution as desired such as easy optimization of the axial power distribution, high flexibility, and flexible flattening of the power distribution, and it requires no special change in view of the design and has a good economical property. (N.H.)

  16. Overall plant concept for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Yamaki, Hideo; Davies, S.M.; Goodman, L.

    1984-01-01

    Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers

  17. Transmutation of plutonium in pebble bed type high temperature reactors

    International Nuclear Information System (INIS)

    Bende, E.E.

    1997-01-01

    The pebble bed type High Temperature Reactor (HTR) has been studied as a uranium-free burner of reactor grade plutonium. In a parametric study, the plutonium loading per pebble as well as the type and size of the coated particles (CPs) have been varied to determine the plutonium consumption, the final plutonium burnup, the k ∞ and the temperature coefficients as a function of burnup. The plutonium loading per pebble is bounded between 1 and 3 gr Pu per pebble. The upper limit is imposed by the maximal allowable fast fluence for the CPs. A higher plutonium loading requires a longer irradiation time to reach a desired burnup, so that the CPs are exposed to a higher fast fluence. The lower limit is determined by the temperature coefficients, which become less negative with increasing moderator-actinide ratio. A burnup of about 600 MWd/kgHM can be reached. With the HTR's high efficiency of 40%, a plutonium supply of 1520 kg/GW e a is achieved. The discharges of plutonium and minor actinides are then 450 and 110 kg/GW e a, respectively. (author)

  18. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Goto, Tadashi; Inoue, Kotaro; Yamakawa, Masanori; Ikeda, Takashi.

    1988-01-01

    Purpose: To promote more positive forcive circulation of primary circuit fluids thereby increase the heat removing amount. Constitution: The primary side of an electromagnetic flow coupler type heat exchanger is opened to the primary fluid of a reactor, while the secondary side is connected with the secondary circuit comprising an air cooler and an electromagnetic pump. Since the secondary circuit stands-by during normal operation, the electromagnetic flow coupler does not operate and does not generate force for flowing primary circuit fluid. If flow due to the external force to the primary circuit fluid should occur in the electromagnetic flow coupler type heat exchanger, an electromagnetic force tending to flow the secondary circuit fluid is exerted oppositely. However the coupler undergoes reaction inertia of the fluid or flowing resistance, to exert in the direction of suppressing the flow, thereby prevent the heat loss. (Yoshihara, H.)

  19. The EPR - technology for the 3rd Millennium

    International Nuclear Information System (INIS)

    Bernstrauch, O.

    2000-01-01

    The Basic Design of the European Pressurized Water Reactor (EPR) was completed 1997 , the Basic Design Optimization Phase 1998 and the Detailed Design Phase will start in the near future. With these milestones, a new generation of PWRs is moving forward. Most of all, this is another story of a successful Franco-German cooperation. It is a rundown of the history of the EPR, before a decision is made to launch the lead-unit construction. The EPR project was launched in 1992 by Nuclear Power International (NPI), a joint company of FRAMATOME and Siemens KWU, supported by EDF and nine German electric utilities. Each step of the development of the EPR was harmonized with the Nuclear Safety Authorities both in France and Germany to reach an early approval. The EPR integrates the latest technological advances, especially in safety and operational aspects and comprises more than 30 years operating experience. Thus, the EPR combines the qualities of its predecessors, the French N4 and the German Konvoi. Presently, Siemens KWU and FRAMATOME are preparing the detailed design phase and the following construction and commissioning phase. The decision to build an EPR is not yet made either by the German electric utilities or by EdF, but it will be expected within the next months as a strong statement to follow the nuclear way and to ensure the know-how transfer. (author)

  20. Liquid-metal-gas heat exchanger for HTGR type reactors

    International Nuclear Information System (INIS)

    Werth, G.

    1980-01-01

    The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL) [de

  1. EPR paradox revisited

    Energy Technology Data Exchange (ETDEWEB)

    Klippert, R. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil)

    1999-07-01

    In a seminal paper from 1935 Einstein, Podolsky and Rosen produced one of the most powerful weapon against the unpredictability of the world ensured by quantum mechanics. The recent production of entangled states, with all its possible future applications in quantum computation, re-open the possibility of testing EPR states on physical grounds. The present intends to present a challenge to the wedding of classical (special) relativity with quantum mechanics, the so called relativistic quantum mechanics. Making use of the same apparatus devised in EPR, it is shown that non local quantum states are incompatible with either their possibility of being measured or else with Lorentz invariance (or even with both). (author)

  2. Development of toroid-type HTS DC reactor series for HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2015-11-15

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  3. Development of toroid-type HTS DC reactor series for HVDC system

    International Nuclear Information System (INIS)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-01-01

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  4. New Generation of self-calibrated SS/EPR dosimeters: Alanine/EPR dosimeters

    International Nuclear Information System (INIS)

    Yordanov, N.D.; Gancheva, V.

    1999-01-01

    A new type of solid state/EPR dosimeters is described. Principally, it contains radiation sensitive diamagnetic material, some quantity of EPR active, but radiation insensitive, substance (for example Mn 2+ /MgO) and a binding material. In the present case alanine is used as a radiation sensitive substance. With this dosimeter, the EPR spectra of alanine and Mn 2+ are simultaneously recorded and the calibration graph represents the ratio of alanine versus Mn 2+ EPR signal intensity as a function of absorbed dose. In this way the reproducibility of the results is expected to be improved significantly including their intercomparison among different laboratories. Homogeneity of the prepared dosimeters and their behaviour (fading of EPR signals with time, influence of different meteorological conditions) show satisfactory reproducibility and stability with time. Because two different EPR active samples are recorded simultaneously, the influence of some instrument setting parameters (microwave power, modulation amplitude and modulation frequency) on the ratio I alanine /I Mn is also investigated. (author)

  5. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    Jaouen, C.; Beroux, P.

    2012-01-01

    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)

  6. EPR spectra of some irradiated polycrystalline perrhenate

    International Nuclear Information System (INIS)

    Zaitseva, N.G.; Constantinescu, M.; Georgescu, R.; Constantinescu, O.

    1978-10-01

    An EPR study of the paramagnetic centers formed by γ, electron and neutron irradiation of the NaReO 4 and KReO 4 was made. In the EPR spectra of the powder samples irradiated γ, with electrons and neutrons, the presence of three types of paramagnetic centers was observed. From the EPR parameters, the centers were attributed to the ReOsub(4)sup(.), ReOsub(3)sup(.) and ReOsub(2)sup(.) radicals respectively. The lower intensity of the spectra observed by KReO 4 samples irradiation showed a higher radioresistance of the KReO 4 than that of NaReO 4 . A radiolitical scheme taking into account the paramagnetic centers formation was proposed. (author)

  7. EPR of impurity ions in disordered solids

    International Nuclear Information System (INIS)

    Kliava, J.

    1986-01-01

    The state of the art in the EPR spectroscopy of disordered solids is reviewed and theoretical aspects of the EPR shape in disordered systems are discussed. Emphasis is placed on the concept of the joint probability density of the spin Hamiltonian parameters. A survey of experimental data is provided on distributions of spin Hamiltonian parametes obtained using computer simulation techniques. A quantitative information is given on the short-range ordering in disordered materials available from EPR studies. A procedure of extracting such type of data which consists in a transformation from the distribution of the spin Hamiltonian parameters to that of atomic coordinates in the surrounding of a paramagnetic center is outlined. Numerical estimates of the degree of continuous disorder are reviewed

  8. Method of stopping operation of PWR type reactor

    International Nuclear Information System (INIS)

    Ueno, Takashi; Tsuge, Ayao; Kawanishi, Yasuhira; Onimura, Kichiro; Kadokami, Akira.

    1989-01-01

    In PWR type reactors after long period of l00 % power operation, since boiling is caused in heat conduction pipes and water is depleted within the intergranular corrosion fracture face in the crevis portion to result in a dry-out state, impregnation and concentration of corrosion inhibitors into the intergranular corrosion fracture face are insufficient. In view of the above, the corrosion inhibitor at a high concentration is impregnated into the intergranular corrosion fracture face by keeping to inject the corrosion inhibitor from l00 % thermal power load by way of the thermal power reduction to the zero power state upon operatioin shutdown. That is, if the thermal power is reduced to or near the 0 power upon reactor shutdown, feedwater in the crevis portion is put to subcooled state, by which the steam present in the intergranular corrosion fracture face are condensated and the corrosion inhibitor at high concentration impregnated into the crevis portion are penetrated into the intergranular corrosion fracture face. (K.M.)

  9. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  10. EPR Dosimetry in Irradiated Fingernails

    International Nuclear Information System (INIS)

    Spinella, M.R.; Dubner, D.L.; Bof, E.

    2010-01-01

    The Electron Paramagnetic Resonance (EPR) is being transformed in a complementary tool of biologically-based methods for evaluation of dose after accidental radiation exposure. Many efforts are being carried out in laboratories to evaluate the performance of different materials for its use in EPR doses measurements and for improving the current methods for spectrum analysis and calibration curves determinations. In our country the EPR techniques have been used in different areas with dosimetric (alanine) and non dosimetric purposes. Now we are performing the first studies to obtain properly dose response curves to be used for accidental dose assessments through irradiated fingernails. It is by now well known that the fingernails present two types of signals, a background one (BKS), originated in elastic and inelastic mechanical deformations and the radio induced one (RIS), object of interest (I). In this work we will present some of the previous studies performed to characterize the fingernail samples and we analyse the additive dose method for data obtained employing the technique of the substraction of the spectrum recorded at two different microwave powers in order to reduce the BKS signal. Fingernail samples collected from different donors were treated by soaking in water during 10 min and 5 min drying on paper towel and the BKS signals were studied previously its irradiation. The statistical analysis (R statistics) show a distribution with a Standard Deviation of 24% respects to its media. During these studies we also conserved in freezer for more than 6 months irradiated fingernails that, were periodically measured and the statistical analysis of the peak to peak amplitude show a normal distribution through the Quantile correlation test with a SD 11% respected to its median. (authors)

  11. EPR: the nuclear impasse

    International Nuclear Information System (INIS)

    Marillier, F.

    2008-01-01

    The questions relative to the climatic change constitute crucial challenges for the next ten years. In this context the author aims to show how the EPR project illustrates the nuclear french ''autism''. He presents and analyzes the international and environmental impacts of this obsolete technology, as a project useless and dangerous. (A.L.B.)

  12. Leak detector for a steam generator in FBR type reactors

    International Nuclear Information System (INIS)

    Miyaji, Nobuyoshi.

    1979-01-01

    Purpose: To facilitate maintenance for liquid leak detectors such as exchange of nickel membrane sensors during operation in a sodium-cooled fbr type reactor. Constitution: A pipeway capable of supplying a cover gas such as argon into the cylinder of a hydrogen detector containing a nickel membrane sensor is provided in a liquid leak detector constituting a part of a by-pass loop. The pipeway is also adapted to be evacuated. A pipeway and a small sodium tank for drain use are provided on the side of the by-pass loop near valves. Then, after closing the inlet and outlet valves to disconnect the by-pass loop from the sodium main pipeway, the cover gas is supplied to drive liquid sodium to the drain tank. After the drain of the liquid sodium, the sensor can be replaced. (Ikeda, J.)

  13. Gamma spectrum measurement in a swimming-pool-type reactor

    International Nuclear Information System (INIS)

    Pla, E.

    1969-01-01

    After recalling the various modes of interaction of gamma rays with matter, the authors describe the design of a spectrometer for gamma energies of between 0.3 and 10 MeV. This spectrometer makes use of the Compton and pair-production effects without eliminating them. The collimator, the crystals and the electronics have been studied in detail and are described in their final form. The problem of calibrating the apparatus is then considered ; numerous graphs are given. The sensitivity of the spectrometer for different energies is determined mainly for the 'Compton effect' group. Finally, in the last part of the report, are given results of an experimental measurement of the gamma spectrum of a swimming-pool type reactor with new elements. (author) [fr

  14. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.; Venescu, R.

    2013-01-01

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  15. IRT-type research reactor physical calculation methodology

    International Nuclear Information System (INIS)

    Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

    1990-01-01

    In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

  16. Consideration of BORAX-type reactivity accidents applied to research reactors

    International Nuclear Information System (INIS)

    Couturier, Jean; Meignen, Renaud; Bourgois, Thierry; Biaut, Guillaume; Mireau, Jean-Pierre; Natta, Marc

    2011-01-01

    Most of the research reactors discussed in this document are pool-type reactors in which the reactor vessel and some of the reactor coolant systems are located in a pool of water. These reactors generally use fuel in plate assemblies formed by a compact layer of uranium (or U 3 Si 2 ) and aluminium particles, sandwiched between two thin layers of aluminium serving as cladding. The fuel melting process begins at 660 deg. C when the aluminium melts, while the uranium (or U 3 Si 2 ) particles may remain solid. The accident that occurred in the American SL-1 reactor in 1961, together with tests carried out in the United States as of 1954 in the BORAX-1 reactor and then, in 1962, in the SPERT-1 reactor, showed that a sudden substantial addition of reactivity in this type of reactor could lead to explosive mechanisms caused by degradation, or even fast meltdown, of part of the reactor core. This is what is known as a 'BORAX-type' accident. The aim of this document is first to briefly recall the circumstances of the SL-1 reactor accident, the lessons learned, how this operational feedback has been factored into the design of various research reactors around the world and, second, to describe the approach taken by France with regard to this type of accident and how, led by IRSN, this approach has evolved in the last decade. (authors)

  17. Task types and error types involved in the human-related unplanned reactor trip events

    International Nuclear Information System (INIS)

    Kim, Jae Whan; Park, Jin Kyun

    2008-01-01

    In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1%), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed

  18. Task types and error types involved in the human-related unplanned reactor trip events

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Park, Jin Kyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1%), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.

  19. Can EPR non-locality be geometrical?

    International Nuclear Information System (INIS)

    Ne'eman, Y.

    1995-01-01

    The presence in Quantum Mechanics of non-local correlations is one of the two fundamentally non-intuitive features of that theory. The non-local correlations themselves fall into two classes: EPR and Geometrical. The non-local characteristics of the geometrical type are well-understood and are not suspected of possibly generating acausal features, such as faster-than-light propagation of information. This has especially become true since the emergence of a geometrical treatment for the relevant gauge theories, i.e. Fiber Bundle geometry, in which the quantum non-localities are seen to correspond to pure homotopy considerations. This aspect is reviewed in section 2. Contrary-wise, from its very conception, the EPR situation was felt to be paradoxical. It has been suggested that the non-local features of EPR might also derive from geometrical considerations, like all other non-local characteristics of QM. In[7], one of the authors was able to point out several plausibility arguments for this thesis, emphasizing in particular similarities between the non-local correlations provided by any gauge field theory and those required by the preservation of the quantum numbers of the original EPR state-vector, throughout its spatially-extended mode. The derivation was, however, somewhat incomplete, especially because of the apparent difference between, on the one hand, the closed spatial loops arising in the analysis of the geometrical non-localities, from Aharonov-Bohm and Berry phases to magnetic monopoles and instantons, and on the other hand, in the EPR case, the open line drawn by the positions of the two moving decay products of the disintegrating particle. In what follows, the authors endeavor to remove this obstacle and show that as in all other QM non-localities, EPR is somehow related to closed loops, almost involving homotopy considerations. They develop this view in section 3

  20. Liquid-poison type power controlling device for nuclear reactor

    International Nuclear Information System (INIS)

    Horiuchi, Tetsuo; Yamanari, Shozo; Sugisaki, Toshihiko; Goto, Hiroshi.

    1981-01-01

    Purpose: To improve the safety and the operability of a nuclear reactor by adjusting the density of liquid poison. Constitution: The thermal expansion follow-up failure between cladding and a pellet upon abrupt and local variations of the power is avoided by adjusting the density of liquid poison during ordinary operation in combination with a high density liquid poison tank and a filter and smoothly controlling the reactor power through a pipe installed in the reactor core. The high density liquid poison is abruptly charged in to the reactor core under relatively low pressure through the tube installed in the reactor core at the time of control rod insertion failure in an accident, thereby effectively shutting down the reactor and improving the safety and the operability of the reactor. (Yoshihara, H.)

  1. EPR by Areva. The path of greatest certainty

    International Nuclear Information System (INIS)

    2008-01-01

    AREVA's Evolutionary Power Reactor (EPR) is the first Generation III+ reactor design currently being built to answer the world's growing demand for clean and reliable electricity generation. Already under construction in Finland, France and China, the EPR is also being considered by America, United Kingdom, South Africa and other countries for the development of their nuclear fleet. The EPR is now clearly destined to become the mainstay of standardized, efficient reactor fleets around the globe. AREVA's EPR incorporates unbeatable know-how provided by an uninterrupted track record of reactor building activities and backed by decades of feedback experience from operating PWRs, including the most recent. The EPR is a Franco-German initiative which benefited from the stringent scrutiny of safety authorities from both countries, at each stage of the project. The EPR has already secured construction licenses from two of the world's most demanding safety authorities in France and Finland and is currently in line for a design certification and a combined construction and operating license (COL) in the USA. It is also taking part in the licensing process recently launched in the United Kingdom. Europe's leading utilities have granted the EPR their approval under the 'European Utilities Requirements' and have further expressed individual interest in the design and performance of the EPR for their businesses. AREVA is the only Gen III+ reactor constructor in the world with ongoing building experience. To date, AREVA is the only vendor who has the necessary field experience that future customers can benefit: - Detailed design completed; - Experience feedback from 87 PWR; - 3 projects going on; - Continuous PWR experience in design and construction. Close to 100% of the EPR primary circuit heavy components are sourced directly from AREVA's integrated plants. Engineering, manufacturing, services and fuel cycle management are totally integrated and mastered by AREVA. From its

  2. EPR by Areva. The path of greatest certainty

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-07-01

    AREVA's Evolutionary Power Reactor (EPR) is the first Generation III+ reactor design currently being built to answer the world's growing demand for clean and reliable electricity generation. Already under construction in Finland, France and China, the EPR is also being considered by America, United Kingdom, South Africa and other countries for the development of their nuclear fleet. The EPR is now clearly destined to become the mainstay of standardized, efficient reactor fleets around the globe. AREVA's EPR incorporates unbeatable know-how provided by an uninterrupted track record of reactor building activities and backed by decades of feedback experience from operating PWRs, including the most recent. The EPR is a Franco-German initiative which benefited from the stringent scrutiny of safety authorities from both countries, at each stage of the project. The EPR has already secured construction licenses from two of the world's most demanding safety authorities in France and Finland and is currently in line for a design certification and a combined construction and operating license (COL) in the USA. It is also taking part in the licensing process recently launched in the United Kingdom. Europe's leading utilities have granted the EPR their approval under the 'European Utilities Requirements' and have further expressed individual interest in the design and performance of the EPR for their businesses. AREVA is the only Gen III+ reactor constructor in the world with ongoing building experience. To date, AREVA is the only vendor who has the necessary field experience that future customers can benefit: - Detailed design completed; - Experience feedback from 87 PWR; - 3 projects going on; - Continuous PWR experience in design and construction. Close to 100% of the EPR primary circuit heavy components are sourced directly from AREVA's integrated plants. Engineering, manufacturing, services and fuel cycle management are totally

  3. Safety classification of systems, structures, and components for pool-type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-08-15

    Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

  4. Development of toroid-type HTS DC reactor series for HVDC system

    Science.gov (United States)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-11-01

    This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  5. EPR becomes reality at Finland's Olkiluoto 3

    International Nuclear Information System (INIS)

    Gueldner, R.; Giese, U.

    2005-01-01

    The EPR is a third-generation pressurized water reactor (PWR). Its development was started in 1992 by Framatome and Siemens within a Franco-German partnership. Since 2001 this work has been continued by Framatome ANP, which was formed when the two companies merged their nuclear businesses. The French company AREVA, world market leader in nuclear technology, holds a 66% share in Framatome ANP, with Siemens owning 34%. From the very start, development of the EPR was focused on improving plant safety and economics even further. The new reactor development was jointly financed together with the leading power utilities of both countries. The first steps towards realization of an EPR nuclear power plant were taken at Olkiluoto, Finland in 2004, consisting of initial preparation of the construction site. By mid-February 2005 the local municipality - Eurajoki - had issued a construction permit, and the Finnish Government a construction license pursuant to the Finnish Nuclear Energy Act. This had been preceded by a preliminary safety assessment prepared by the Finnish Radiation and Nuclear Safety Authority (STUK) for the Finnish Ministry of Trade and Industry in which STUK verified that it did not see any safety-related issues opposing issuance of the nuclear construction license. STUK emphasized that the evolutionary design of the EPR had been further improved by AREVA compared to the previous product lines. Concreting work began this spring and the unit will start commercial operation in 2009. Construction of an EPR has also been given the political go-ahead in France. According to the utility Electricite de France (EDF) the new reactor will be built as a forerunner of a later series at the site of Flamanville in Normandy. Construction is scheduled to begin in 2007. An EPR nuclear power plant has a rated electric capacity of around 1600 MW, depending on specific site conditions. Being the product of intense bilateral cooperation the EPR combines the technological

  6. Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

    International Nuclear Information System (INIS)

    Goswami, S.

    1983-01-01

    The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

  7. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  8. In-service inspection of pool type research reactors

    International Nuclear Information System (INIS)

    Rajamani, K.

    2002-01-01

    In the case of Apsara Reactor, it has been proposed to carry out major modifications in the near future. It is planned to modify the core suitably with a heavy water reflector tank to demonstrate the Multiple Purpose Research Reactor concept. The core structure will be a stationary one and will be located at the 'B' position of the pool. The modified reactor will be operated at 1 MW power level. Suitable methodologies are evolved for carrying out a planned ISI for this modified reactor

  9. Method of cooling a pressure tube type reactor

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro.

    1983-01-01

    Purpose: To improve the operation efficiency of a nuclear reactor by carrying out cooling depending on the power distribution in the reactor core. Constitution: Reactor core channels are divided into a plurality of channel groups depending on the reactor power, and a water drum and a pump are disposed to each of the channel groups so as to increase the amount of coolants in response to the magnitude of the power from each of the channel groups. In this way, the minimum limiting power ratio can be increased. (Seki, T.)

  10. EPR, kvantemekanik og Bohr

    OpenAIRE

    Nielsen, Morten Klockmann

    2007-01-01

    Dette projekt omhandler området hvor filosofi og fysik smelter sammen. Kvantemekanikkens tilblivelse fik en hård medfart hvilket diskussionerne mellem især Albert Einstein og Niels Bohr vidner om. De var hovedpersoner i striden om hvordan kvantemekanikken skulle fortolkes, og diskussionen kulminerede i 1935 hvor Einstein sammen med kollegerne Podolsky og Rosen offentliggjorde en artikel med titlen “Can Quantum-Mechanical Description of Physical Reality Be Considered Complete?” (EPR-artiklen)....

  11. Seismic responses of a pool-type fast reactor with different core support designs

    International Nuclear Information System (INIS)

    Wu, Ting-shu; Seidensticker, R.W.

    1989-01-01

    In designing the core support system for a pool-type fast reactor, there are many issues which must be considered in order to achieve an optimum and balanced design. These issues include safety, reliability, as well as costs. Several design options are possible to support the reactor core. Different core support options yield different frequency ranges and responses. Seismic responses of a large pool-type fast reactor incorporated with different core support designs have been investigated. 4 refs., 3 figs

  12. Problems of control of WWER-type pressurized water reactors (PWR's)

    International Nuclear Information System (INIS)

    Drab, F.; Grof, V.

    1978-01-01

    The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

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

    International Nuclear Information System (INIS)

    Marcon, M.; Faugere, J.L.; Genthon, J.P.; Maillot, R.

    1977-01-01

    Nuclear fuel tests for the fast neutron reactor type have been carried out at the Osiris reactor: thermal study of (U,Pu)O 2 oxide by measurement with thermocouples in the core of the fuel pellet; study of the effects of power cycling on nuclear fuel; study of the mechanical interactions between oxide and cladding by measurement of the cladding deformation during irradiation [fr

  14. EPR spectroscopy of spices

    Directory of Open Access Journals (Sweden)

    R. T. Тimakova

    2016-01-01

    Full Text Available From 01 January 2017 you enter the interstate standard GOST 33271-2015 “Dry Spices, herbs and vegetable seasonings. Manual exposure in order to combat pathogens and other microorganisms” which States that the absorbed dose of radiation to the spices should be from 3 to 30 kGy. The study found that before the introduction of permissive legislative framework in the consumer market of Russia there are irradiated food products (chili, ground chili, ground spicy chili, black pepper. For radiation monitoring of food safety, we used the method of electron paramagnetic resonance (EPR, which allows quickly and with a high degree of reliability to establish the fact of irradiation. It is established that all samples of spices irradiated with dose of 12 kGy (technology radappertization gave typical spectra of the signals established by the method of electron paramagnetic resonance in the domestic EPR spectrometer, the intensity, amplitude and peak width of the EPR signal of samples of spices with the increase of irradiation dose increases. It is proven that repeated exposure no effect accumulation. Integration with 2017 Russia in the global practi ce of using radiation technologies of processing of food products and food raw materials with the purpose of extending shelf life confirms the need for a data Bank on the radiation sensitivity of various food products to determine the optimal doses and the eff ect of radiation doses on the shelf life and quality of products.

  15. Electron paramagnetic resonance (EPR) biodosimetry

    International Nuclear Information System (INIS)

    Desrosiers, Marc; Schauer, David A.

    2001-01-01

    Radiation-induced electron paramagnetic resonance (EPR) signals were first reported by Gordy et al. [Proc. Natl. Acad. Sci. USA 41 (1955) 983]. The application of EPR spectroscopy to ionizing radiation dosimetry was later proposed by Brady et al. [Health Phys. 15 (1968) 43]. Since that time EPR dosimetry has been applied to accident and epidemiologic dose reconstruction, radiation therapy, food irradiation, quality assurance programs and archaeological dating. Materials that have been studied include bone, tooth enamel, alanine and quartz. This review paper presents the fundamentals and applications of EPR biodosimetry. Detailed information regarding sample collection and preparation, EPR measurements, dose reconstruction, and data analysis and interpretation will be reviewed for tooth enamel. Examples of EPR biodosimetry application in accidental overexposures, radiopharmaceutical dose assessment and retrospective epidemiologic studies will also be presented

  16. Opinion on the examination of the control-command general architecture of the Flamanville 3 EPR reactor and of its associated platforms

    International Nuclear Information System (INIS)

    2009-06-01

    This report first recalls the different technical and hardware elements and components discussed by the ISRN (Nuclear Radioprotection and Safety Institute) in front of the Expert Permanent Group for nuclear reactors. These elements and components are concerning the control-command architecture as a whole, and the technical ability of networks, computers, and robots. It outlines the main principles which the control-and-command must comply with: compliance of each component of the retained technological solution with the design requirements corresponding to the safety classification, robustness of the whole control-command architecture. Then, it comments and sometimes criticizes the status of the various components of the control-command platforms with respect to this compliance. Eight recommendations are given for the SPPA T2000 control-command platform and for the robustness of the control-command architecture

  17. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  18. Review of the general atomic experimental fusion power reactor initial conceptual design

    International Nuclear Information System (INIS)

    Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

    1976-01-01

    The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

  19. EPR spectral investigation of radiation-induced radicals of gallic acid

    International Nuclear Information System (INIS)

    Tuner, Hasan

    2017-01-01

    In the present work, spectroscopic features of the radiation-induced radicals of gallic acid compounds were investigated using electron paramagnetic resonance (EPR) spectroscopy. While un-irradiated samples presented no EPR signal, irradiated samples exhibited an EPR spectrum consisting of an intense resonance line at the center and weak lines on both sides. Detailed microwave saturation investigations were carried out to determine the origin of the experimental EPR lines. It is concluded that the two side lines of the triplet satellite originate from forbidden ''spin-flip'' transitions. The spectroscopic and structural features of the radiation-induced radicals were determined using EPR spectrum fittings. The experimental EPR spectra of the two gallic acid compounds were consistent with the calculated EPR spectroscopic features of the proposed radicals. It is concluded that the most probable radicals are the cyclohexadienyl-type, O(OH) 2 C 6 H 2 COOH radicals for both compounds. (orig.)

  20. EPR spectral investigation of radiation-induced radicals of gallic acid.

    Science.gov (United States)

    Tuner, Hasan

    2017-11-01

    In the present work, spectroscopic features of the radiation-induced radicals of gallic acid compounds were investigated using electron paramagnetic resonance (EPR) spectroscopy. While un-irradiated samples presented no EPR signal, irradiated samples exhibited an EPR spectrum consisting of an intense resonance line at the center and weak lines on both sides. Detailed microwave saturation investigations were carried out to determine the origin of the experimental EPR lines. It is concluded that the two side lines of the triplet satellite originate from forbidden "spin-flip" transitions. The spectroscopic and structural features of the radiation-induced radicals were determined using EPR spectrum fittings. The experimental EPR spectra of the two gallic acid compounds were consistent with the calculated EPR spectroscopic features of the proposed radicals. It is concluded that the most probable radicals are the cyclohexadienyl-type, [Formula: see text] radicals for both compounds.

  1. EPR spectroscopy at DNP conditions

    International Nuclear Information System (INIS)

    Heckmann, J.; Goertz, St.; Meyer, W.; Radtke, E.; Reicherz, G.

    2004-01-01

    In terms of dynamic nuclear polarization (DNP) studies and systematic target material research it is crucial to know the EPR lineshape of the DNP relevant paramagnetic centers. Therefore in Bochum an EPR spectrometer has been implemented into the 4 He evaporation DNP facility in order to perform EPR studies at DNP conditions (B=2.5 T, T=1 K). The spectrometer hardware and performance as well as first results are presented

  2. On the slimeless water operation in the RBMK type reactors

    International Nuclear Information System (INIS)

    Margulova, T.Kh.; Mamet, V.A.; Nikitina, I.S.; Karakhanyan, L.N.

    1988-01-01

    Water chemistry conditions of the operating RBMK-1000 and RBMK-1500 units are analysed. Inevitability of iron oxide deposits in RBMK-1000 and particularly in RBMK-1500 reactors is demonstrated. Organization of a new slimeless correcting water chemistry for RBMK-1000 and RBMK-1500 reactors is recommended

  3. Auxiliary water supply device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

  4. Fuel assemblies for use in BWR type reactors

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1987-01-01

    Purpose: To moderate the peak configuration of the burnup degree change curve for the infinite multiplication factor by applying an improvement to the arrangement of fuel rods. Constitution: In a fuel assembly for a BWR type reactor comprising a plurality of fuel rods and water rods arranged in a square lattice, fuel rods containing burnable poisons are arranged at four corners at the second and the third layers from the outside of the square lattice arrangement. Among them, the Cd poison effect in the burnable poison incorporated fuel rods disposed at the second layer is somewhat greater at the initial burning stage and then rapidly decreased along with burning. While on the other hand, the poison effect of the burnable poison-incorporated fuel rods at the third layer is smaller than that at the second layer at the initial burning stage and the reduction in the poison effect due to burning is somewhat more moderate. Since these fuel rods are in adjacent with each other, they interfere to each other and also provide an effect of moderating the burning of the burnable poisons. (Takahashi, M.)

  5. Deploying the advanced U.S. EPR

    International Nuclear Information System (INIS)

    Hagan, C.

    2007-01-01

    As electricity demand, clean air concerns and energy prices increase in the U.S., interest in new nuclear reactors is also on the rise. These factors - along with decades of outstanding performance of the nation's existing nuclear fleet - have contributed to a favorable market for new nuclear power in the U.S. Although no U.S. company has declared it actually will buy and build new reactors, many have actively expressed their interest through investment and regulatory interaction. The timing is also right for new reactors in terms of public perception, government policy and acceptance of the investment community. Amid this environment, the U.S. commercial power industry recently has begun pursuing new nuclear generation in earnest. AREVA, whose ongoing global market analyses indicated the need for additional baseload nuclear generation in the U.S., chose to license its EPR design to meet that need. This article will discuss the American market landscape for new nuclear and how AREVA is approaching it, present an overview of the company's U.S. EPR technology, and provide an update on the regulatory status of the design and deployment schedule. (orig.)

  6. Quantitative EPR A Practitioners Guide

    CERN Document Server

    Eaton, Gareth R; Barr, David P; Weber, Ralph T

    2010-01-01

    This is the first comprehensive yet practical guide for people who perform quantitative EPR measurements. No existing book provides this level of practical guidance to ensure the successful use of EPR. There is a growing need in both industrial and academic research to provide meaningful and accurate quantitative EPR results. This text discusses the various sample, instrument and software related aspects required for EPR quantitation. Specific topics include: choosing a reference standard, resonator considerations (Q, B1, Bm), power saturation characteristics, sample positioning, and finally, putting all the factors together to obtain an accurate spin concentration of a sample.

  7. Active or passive systems? The EPR approach

    International Nuclear Information System (INIS)

    Bonhomme, N.; Py, J.P.

    1996-01-01

    In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs

  8. Active or passive systems? The EPR approach

    Energy Technology Data Exchange (ETDEWEB)

    Bonhomme, N [Nuclear Power International, Cedex (France); Py, J P [FRAMATOME, Cedex (France)

    1996-12-01

    In attempting to review how EPR is contemplated to meet requirements applicable to future nuclear power plants, the authors indicate where they see the markets and the corresponding unit sizes for the EPR which is a generic key factor for competitiveness. There are no reason in industrialized countries, other than USA (where the investment and amortizing practices under control by Public Utility Commission are quite particular), not to build future plants in the 1000 to 1500 MWe range. Standardization, which has been actively applied all along the French program and for the Konvoi plants, does not prevent evolution and allows to concentrate large engineering effort in smooth realization of plants and achieve actual construction and commissioning without significant delays. In order to contribute to public trust renewal, a next generation of power reactors should be fundamentally less likely to incur serious accidents. To reach this goal the best of passive and active systems must be considered without forgetting that the most important source of knowledge is construction and operating experience. Criteria to assess passive systems investigated for possible implementation in the EPR, such as simplicity of design, impact on plant operation, safety and cost, are discussed. Examples of the principal passive systems investigated are described and reasons why they have been dropped after screening through the criteria are given. (author). 11 figs.

  9. Which Parts of a Clinical Process EPR Needs Special Configuration

    DEFF Research Database (Denmark)

    Barlach, Anders; Simonsen, Jesper

    2007-01-01

    Subject: Which parts of an electronic patient record (EPR) can initially form a stable standard solution to be used by all clinicians? And which parts of an EPR can we predict needs initial as well as on-going re-configuration to meet the needs from diverse medical specialties. Purpose: To analyze...... which screen types in a clinical process that can be standard configured and which are subject to initial as well as on-going re-configuration. Methods and results: A pilot-project implementing a fully functional clinical process EPR was configured and used at a neurological ward, replacing all paper...... records 24/7. The analysis characterizes the different types of screens, a total of 243 included in the EPR solution. All screens have been extracted from the application and analyzed for changes – in total 222 changes. Discussion and conclusion: Most screens (87%) are very stable. Few (13%) are subjected...

  10. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  11. The Einstein-Podolsky-Rosen Paradox and Entanglement 1: Signatures of EPR correlations for continuous variables

    OpenAIRE

    Reid, M. D.

    2001-01-01

    A generalization of the 1935 Einstein-Podolsky-Rosen (EPR) argument for measurements with continuous variable outcomes is presented to establish criteria for the demonstration of the EPR paradox, for situations where the correlation between spatially separated subsystems is not perfect. Two types of criteria for EPR correlations are determined. The first type are based on measurements of the variances of conditional probability distributions and are necessary to reflect directly the situation...

  12. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  13. HTGR type reactors in West Germany. Realizations and prospects

    International Nuclear Information System (INIS)

    Dauenert, U.

    1978-01-01

    The framework within which the research studies on high temperature reactors have been pursued in West Germany since 1960 is recalled. The principles guiding the present policy of the country in this domain are given: choice of a single technical conception that be applied both to reactors generating electricity and reactors producing high temperature heat for industrial processes such as coal gasification; to group the technical and industrial potentials of West Germany in this domain; financial and technical participation of electricity producers in the expected realizations; international cooperation. In this technique, West Germany is at present among the most advanced nations with the realization of a prototype 300 MWe reactor, financed by the electricity producers and a contribution of government [fr

  14. Startup method for natural convection type nuclear reactor

    International Nuclear Information System (INIS)

    Utsuno, Hideaki.

    1993-01-01

    In a nuclear reactor started by natural convection, no sufficient stability margin can be ensured upon start up. Then, in the present invention, a deaerating operation is conducted before start-up of the reactor, then control rods are withdrawn after conducting the deaerating operation and temperature and pressure are raised by nuclear heating, to obtain a rated power. As a result, reactor power and subcooling at the inlet of the reactor core are within a range of lower than a geysering forming region, thereby enabling to prevent occurence of geysering inherent to the start-up of operation in a natural convection state, shorten the start-up time, as well as remove oxygen dissolved in coolants. (N.H.)

  15. Seismic response of a block-type nuclear reactor core

    International Nuclear Information System (INIS)

    Dove, R.C.; Bennett, J.G.; Merson, J.L.

    1976-05-01

    An analytical model is developed to predict seismic response of large gas-cooled reactor cores. The model is used to investigate scaling laws involved in the design of physical models of such cores, and to make parameter studies

  16. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-01-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  17. Radiological characterization for small type light water reactor

    International Nuclear Information System (INIS)

    Tanaka, Ken-ichi; Ichige, Hideaki; Tanabe, Hidenori

    2011-01-01

    In order to plan a decommissioning, amount investigation of waste materials and residual radioactivity inventory evaluation must be performed at the first stage of preparatory tasks. These tasks are called radiological characterization. Reliable information from radiological characterization is crucial for specification of decommissioning plan. With the information, we can perform radiological safety analysis and optimize decommissioning scenario. Japan Atomic Power Company (JAPC) has already started preparatory tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1) that is the first commercial Small Type Light Water Reactor in Japan. To obtain reliable information about residual radioactivity inventory, we improved radioactivity inventory evaluation procedure. The procedure consists of neutron flux distribution calculation and radioactivity distribution calculation. We need a better understanding about characteristics of neutron transport phenomena in order to obtain reliable neutron flux distribution. Neutron flux was measured in Primary Containment Vessel (PCV) at 30 locations using activation foils. We chose locations where characteristic phenomena can be observed. Three dimensional (3D) neutron flux calculation was also performed to simulate continuous changes of neutron flux distribution. By assessing both the measured values and 3D calculation results, we could perform the calculation that simulates the phenomena well. We got knowledge about how to perform an appropriate neutron flux distribution calculation and also became able to calculate a reliable neutron flux distribution. Using the neutron flux distribution, we can estimate a reliable radioactivity distribution. We applied network-parallel-computing method to the estimation. And further we developed 'flux level approximation method' which use linear or parabola fitting method to estimation. Using these new methods, radioactivity by neutron irradiation, which is radioisotope formation, was calculated at

  18. Emergency core cooling system in BWR type reactors

    International Nuclear Information System (INIS)

    Takizawa, Yoji

    1981-01-01

    Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

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

    International Nuclear Information System (INIS)

    Niwa, Hajime; Aoto, Kazumi; Morishita, Masaki

    2007-01-01

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

  20. Teichmuller Space Resolution of the EPR Paradox

    Science.gov (United States)

    Winterberg, Friedwardt

    2013-04-01

    The mystery of Newton's action-at-a-distance law of gravity was resolved by Einstein with Riemann's non-Euclidean geometry, which permitted the explanation of the departure from Newton's law for the motion of Mercury. It is here proposed that the similarly mysterious non-local EPR-type quantum correlations may be explained by a Teichmuller space geometry below the Planck length, for which an experiment for its verification is proposed.

  1. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  2. Recycling flow rate control device in BWR type reactor

    International Nuclear Information System (INIS)

    Fujiwara, Tadashi; Koda, Yasushi

    1988-01-01

    Purpose: To reduce the recycling pump speed if the pressure variation width and the variation ratio in the nuclear reactor exceed predetermined values, to thereby avoid the shutdown of the plant. Constitution: There has been proposed a method of monitoring the neutron flux increase thereby avoiding unnecessary plant shutdown, but it involves a problems of reactor scram depending on the state of the plant and the set values. In view of the above, in the plant using internal pumps put under the thyristor control and having high response to recycling flow rate, the reactor pressure is monitored and the speed of the internal pump is rapidly reduced when the pressure variation width and variation ratio exceed predetermined values to reduce the reactor power and avoid the plant shutdown. This can reduce the possibility of unnecessary power reduction due to neutron flux noises or the possibility of plant shutdown under low power conditions. Further, since the reactor operation can be continued without stopping the recycling pump, the operation upon recovery can be made rapid. (Horiuchi, T.)

  3. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  4. EPR of radiation defects in lithium-oxyfluoride glass ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Fedotovs, A; Rogulis, U; Sarakovskis, A; Dimitrocenko, L, E-mail: andris-f@navigator.l [Institute of Solid State Physics, University of Latvia, Kengaraga st. 8, LV-1063, Riga (Latvia)

    2010-11-01

    We studied oxyfluoride composites based on lithium silicate glasses with yttrium fluorides and rare-earth dopants. The electron paramagnetic resonance (EPR) has been used to obtain information about radiation induced defects in these materials. Spectra have been measured before and after X-ray irradiation at room temperature and at liquid nitrogen temperature. Fluoride crystallites within samples were created by means of thermal treatment at specific temperatures. EPR spectra of radiation induced defects in oxyfluoride glass ceramics, in which crystallites have not been yet created, show no explicit hfs interaction of fluorine nuclei. However, in glass ceramics, which already contains fluoride crystallites, the hfs characteristic to fluorine nuclei appears in the EPR spectra. EPR hyperfine structure could be explained within a model of an F-type centre in YF{sub 3} crystalline phase.

  5. EPR of radiation defects in lithium-oxyfluoride glass ceramics

    Science.gov (United States)

    Fedotovs, A.; Rogulis, U.; Sarakovskis, A.; Dimitrocenko, L.

    2010-11-01

    We studied oxyfluoride composites based on lithium silicate glasses with yttrium fluorides and rare-earth dopants. The electron paramagnetic resonance (EPR) has been used to obtain information about radiation induced defects in these materials. Spectra have been measured before and after X-ray irradiation at room temperature and at liquid nitrogen temperature. Fluoride crystallites within samples were created by means of thermal treatment at specific temperatures. EPR spectra of radiation induced defects in oxyfluoride glass ceramics, in which crystallites have not been yet created, show no explicit hfs interaction of fluorine nuclei. However, in glass ceramics, which already contains fluoride crystallites, the hfs characteristic to fluorine nuclei appears in the EPR spectra. EPR hyperfine structure could be explained within a model of an F-type centre in YF3 crystalline phase.

  6. Capital cost evaluation of liquid metal reactor by plant type - comparison of modular type with monolithic type -

    International Nuclear Information System (INIS)

    Mun, K. H.; Seok, S. D.; Song, K. D.; Kim, I. C.

    1999-01-01

    A preliminary economic comparison study was performed for KALIMER(Korea Advanced LIquid MEtal Reactor)between a modular plant type with 8 150MWe modules and a 1200MWe monolithic plant type. In both cases of FOAK (First-Of-A-Kind) Plant and NOAK (Nth-Of-A-Kind) Plant, the result says that the economics of monolithic plant is superior to its modular plant. In case of NOAK plant comparison, however, the cost difference is not significant. It means that modular plant can compete with monolithic plant in capital cost if it makes efforts of cost reduction and technical progress on the assumption that the same type of NOAK plant will be constructed continuously

  7. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    Science.gov (United States)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  8. Electroerosion cutting of low-sized templets from WWER-1000 type reactor vessel

    International Nuclear Information System (INIS)

    Neklyudov, I.M.; Ozhigov, L.S.; Gozhenko, S.V.

    2012-01-01

    The article presents the results of developed method of electroerosion cutting of low-sized templets for the reactor vessel metal composition and structure control in laboratory environment. The article describes the equipment for the remote electroerosive cutting of templets from WWER-1000 type reactor vessel by rigid electrode. The testing results are also shown.

  9. Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

    International Nuclear Information System (INIS)

    Delafosse, Jacques.

    1976-01-01

    French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

  10. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type

    International Nuclear Information System (INIS)

    Bresard, I.; Bonal, J.P.

    2000-01-01

    In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)

  11. The development of the physical conceptions of the FBR type reactors control methods

    International Nuclear Information System (INIS)

    Matveev, V.I.; Ivanov, A.P.

    1984-01-01

    The physical concepts and specific problems of the control elements for LMFBR type reactors are discussed in this paper. Typical temperature coefficient of reactivity, its dependency on reactor power and burnup level are given. The authors give us the most advisable methods of the reactivity coefficient compensation

  12. Digital reactor period meter type of NSSG-7

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The paper presents the idea and electronic circuits of the Digital Reactor Period Meter. The instrument consists of a neutron ionisation chamber, the amplifier logarithming the output chamber current, the circuit taking two samples of the log amplifier output signal and subtracting them, the analog -to -digital dividing circuit and the scaler providing the final information of the reactor period value in seconds and in the digital form. Besides it, the instrument produces the acoustic signal in the case, when the rise-time of neutron flux exceeds the permitted value. The untypical construction of the reactor period meter has been developed to obtain both good measurement accuracy and the resistance against the electromagnetic background pulses interfering with the measuring process. The applied measuring system has been patented.

  13. Identification of process dynamics. Stability monitoring in BWR type reactors

    International Nuclear Information System (INIS)

    Abrahamsson, P.; Hallgren, P.

    1991-06-01

    Identification of process dynamics is used for stability monitoring in nuclear reactors (Boiling Water Reactor). This report treats the problem of estimating a damping factor and a resonance frequency from the neutron flux as measured in the reactor. A new parametric online method for identification is derived and presented, and is shown to meet the requirements of stability monitoring. The technique for estimating the process parameters is based on a recursive lattice filter algorithm. The problem of time varying parameters and offset, as well as offline experiments and signal processing are treated. All parts are implemented in a realtime program, using the language C. In comparison with earlier identifications, the new way of estimating the damping factor is shown to work well. Estimates of both the damping factor and the resonance frequency show a stable and reliable behavior. Future development and improvements are also indicated. (au)

  14. Mathematical game type optimization of powerful fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    To obtain maximum speed of putting into operation fast breeders it is recommended on the initial stage of putting into operation these reactors to apply lower power which needs less fission materials. That is why there is an attempt to find a configuration of a high-power reactor providing maximum power for minimum mass of fission material. This problem has a structure of the mathematical game with two partners of non-zero-order total and is solved by means of specific aids of theory of games. Optimal distribution of fission and breeding materials in a multizone reactor first is determined by solution of competitive game and then, on its base, by solution of the cooperation game. The second problem the solution for which is searched is developed from remark on the fact that a reactor with minimum coefficient of flux heterogenity has a configuration different from the reactor with power coefficient heterogenity. Maximum burn-up of fuel needs minimum heterogenity of the flux coefficient and the highest power level needs minimum coefficient of power heterogenity. That is why it is possible to put a problem of finding of the reactor configuration having both coefficients with minimum value. This problem has a structure of a mathematical game with two partners of non-zero-order total and is solved analogously giving optimal distribution of fuel from the new point of view. In the report is shown that both these solutions are independent which is a result of the aim put in the problem of optimization. (author)

  15. Assessment of the integrity of WWER type reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Procedures are given for the assessment of the residual lifetime of reactor pressure vessels with respect to a sudden failure, the lifetime of vessels with defects disclosed during in-service inspections, and the fatigue or corrosion-mechanical lifetime. Also outlined are the ways of assessing the effects of major degradation mechanisms, i.e. radiation embrittlement, thermal aging, and fatigue damage, including the use of calculated values and experimental examination, by means of surveillance specimens in particular. All results of assessment performed so far indicate that the life of reactor pressure vessels at the Dukovany, Jaslovske Bohunice, and Temelin nuclear power plants is well secured. 7 figs., 3 refs

  16. Study on regimes of nuclear power plants with WWER-type reactors

    International Nuclear Information System (INIS)

    Akkerman, G.; Khampel', R.; Khentshel', G.; Kertsher, F.; Lyuttsov, K.

    1976-01-01

    The problems are considered of optimization of nuclear fuel loading, the peculiarities of the NPP operation at decreased power, and also the problem of stability operation of NPP with WWER type reactors taking into account specific features of these reactors (partial fuel overloads, change in reactor reactivity with power changes). The two particular interconnected problems discussed are: choice of such a sequence of partial rechargings which ensures the minimum cost of the electric power generated, and increasing the reactor operating time by reducing its power output. Besides the technical and economic estimates, much attention is given to analysing the stability of NPP operation

  17. Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

    International Nuclear Information System (INIS)

    Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

    1993-01-01

    Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

  18. The EPR - technology for the 3rd millennium

    International Nuclear Information System (INIS)

    Bernstrauch, O.; Dubois, F.

    2001-01-01

    This paper is a story of a successful Franco-German co-operation: the EPR project. It is a rundown of the history of the EPR, before a decision is made to launch the lead-unit construction. The EPR is expected to progressively replace the existing nuclear power plants when they will reach the end of their service life. It integrates the latest technological advances, especially in safety and operational aspects and comprises more than 30 years operating experience. Thus, the EPR combines the qualities of its predecessors, the French N4 and the German Konvoi. The safety of the EPR was maximized both to prevent hypothetical accidents - even severe ones - and to reduce their consequences (corium spreading area, pre-stressed double containment with local liner, four-train safety systems...). The n+2 strategy (in the event of a problem with one train and even if a second train is undergoing maintenance, the remaining two are sufficient to perform the function) allows to perform maintenance during operation which leads consequently to short outage periods, highest availabilities and economical operation. Moreover, most of the EPR components are the result of mastered evolution. The reactor vessel, as a key element for the reactor service life, is designed to be in service for 60 years, the core is large (241 fuel elements instead of 205 in the N4 or 193 in the Konvoi) and the steam generators have higher efficiency. Along the same line, the core and core barrel design have been modified to allow a reduced uranium enrichment and the increase to 65 Gigawatt-day per metric-ton of the fuel discharge burn-up. Finally, the instrumentation and control systems minimize human error by giving the operators a grace period of at least 30 minutes to make decisions. (authors)

  19. The economic potential of a cassette-type-reactor-installed nuclear ice-breaking container ship

    International Nuclear Information System (INIS)

    Kondo, K.; Takamasa, T.

    2000-01-01

    The design concept of the cassette-type-reactor MRX (Marine Reactor X), being under development in Japan for the nuclear ice-breaker container ship is described. The MRX reactor is the monoblock water-cooled and moderated reactor with passive cooling system of natural circulation. It is shown that application of the reactor being under consideration gives an opportunity to decrease greatly the difference in prices for similar nuclear and diesel ships. Economic estimations for applicability of the nuclear ice-breaker container ship with the MRX reactor in Arctics for transportation of standard containers TEU from Europe to Far East as compared with transportation of the same containers by diesel ships via Suets Canal are made [ru

  20. Power distribution monitoring and control in the RBMK type reactors

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Postnikov, V.V.; Volod'ko, Yu.I.

    1980-01-01

    Considered are the structures of monitoring and control systems for the RBMK-1000 reactor including three main systems with high independence: the control and safety system (CSS); the system for physical control of energy distribution (SPCED) as well as the Scala system for centralized control (SCC). Main functions and peculiarities of each system are discussed. Main attention is paid to new structural solutions and new equipment components used in these systems. Described are the RBMK operation software and routine of energy distribution control in it. It is noted that the set of reactor control and monitoring systems has a hierarchical structure, the first level of which includes analog systems (CSS and SPCED) normalizing and transmitting detector signals to the systems of the second level based on computers and realizing computer data processing, data representation to the operator, automatic (through CSS) control for energy distribution, diagnostics of equipment condition and local safety with provision for existing reserves with respect to crisis and thermal loading of fuel assemblies. The third level includes a power computer carrying out complex physical and optimization calculations and providing interconnections with the external computer of power system. A typical feature of the complex is the provision of local automatic safety of the reactor from erroneous withdrawal of any control rod. The complex is designed for complete automatization of energy distribution control in reactor in steady and transient operation conditions

  1. Superhigh-frequency circuit for the EPR spectrometer with rectifier screening

    International Nuclear Information System (INIS)

    Zhizhchenko, G.A.; Tsvirko, L.V.

    1983-01-01

    The hamodyne SHF circuit of a 3-cm EPR spectrometer with a reflecting resonator is described. The optimum operating mode of SHF-rectifier at a constant phase difference is automatically assured in the circuit. The circuit employs a reflecting p-i-n- attenuator and a SHF-rectifier sereen which simplify the spectrometer tuming. The circuit is used in a miniature EPR radiospectrometer Minsk EPR-6-type

  2. Reactor power automatically controlling method and device for BWR type reactor

    International Nuclear Information System (INIS)

    Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

    1997-01-01

    For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

  3. EDF's (Electricite de France) in service control for GCR type reactor vessels

    International Nuclear Information System (INIS)

    Douillet, M.G.

    1979-01-01

    This paper presents the performance of the data acquisition and processing systems developed by the French EDF for controlling and testing the mechanical properties (thermal stress, deformations, cracks,...) of prestressed concrete vessels for GCR type reactors

  4. EPR of uranium ions

    International Nuclear Information System (INIS)

    Ursu, I.; Lupei, V.

    1984-02-01

    A review of the electron paramagnetic resonance data on the uranium ions is given. After a general account of the electronic structure of the uranium free atoms and ions, the influence of the external fields (magnetic field, crystal fields) is discussed. The main information obtained from EPR studies on the uranium ions in crystals are emphasized: identification of the valence and of the ground electronic state, determination of the structure of the centers, crystal field effects, role of the intermediate coupling and of the J-mixing, role of the covalency, determination of the nuclear spin, maqnetic dipole moment and electric quadrupole moment of the odd isotopes of uranium. These data emphasize the fact that the actinide group has its own identity and this is accutely manifested at the beginning of the 5fsup(n) series encompassed by the uranium ions. (authors)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  6. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Moreira, F.J.

    1984-01-01

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

  7. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1987-01-01

    A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

  8. Concept of magnet systems for LHD-type reactor

    International Nuclear Information System (INIS)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2008-10-01

    Heliotron reactors have attractive features for fusion power plants, such as no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, the major radius of plasma of 14 to 17 m with a central toroidal field of 6 to 4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120 to 140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress can be comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than 1,000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is longer than 150 m that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with small extension of the ITER technology. (author)

  9. Response of Voronezh reactor type to horizontal ground motion

    International Nuclear Information System (INIS)

    Pecinka, L.

    1983-01-01

    For the purposes of vibration monitoring of PWR's the well known 'double pendulum model' has been developed and experimentally verified. It is shown, that this model is possible to use for response calculations of Voronezh reactor pressure vessel and its internals to horizontal ground motion. The equation of motion is given in usual matrix form, the damping matrix is calculated by Rayleigh formula. Driving force is given by vector of ground motion in horizontal direction. For the numerical integration of equation of motion is possible to use following methods - matrix exponential in state space; - modal analysis; - one-step direct integration. For our purposes the last one has been chosen and related computer code TRANS has been developed. The results of calculations are given in the graphically form using generalized angular coordinates and its second derivatives, which describes the displacement or acceleration of reactor pressure vessel to ground and the core barrel to reactor pressure vessel. The driving vector is given in the form of artificially generated accelerogram. (orig./HP)

  10. Method for pre-heating lmfbr type reactors

    International Nuclear Information System (INIS)

    Yokozawa, Atsushi; Kataoka, Hajime.

    1978-01-01

    Purpose: To enable pre-heating for the inside of the reactor container and the inside of the coolant recycling system with no additional facilities. Method: The coolant recycling system is composed of a heat exchanger, a mechanical pump, a check valve, a flow meter or the like and it is connected in series by way of a pipe line to a reactor container. The mechanical pump is used as a gas recycling device upon pre-heating and it is designed so that a blower such as a fan can be replaced for the impeller of the pump. The inside of the reactor container and the inside of the coolant recycling system is at first filled with an inert gas such as for use with cover gas. Then, nuclear fuels are loaded to attain criticality. Simultaneously, the blower is started and the control rods are operated while cooling the nuclear fuel with the inert gas thus to obtain heat required for pre-heating the pipe line or the like from the nuclear fuels. After the completion of the pre-heating, the liquid metal is charged. (Ikeda, J.)

  11. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  12. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  13. Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods

    International Nuclear Information System (INIS)

    Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan

    2015-01-01

    Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied

  14. Conceptual design of control rod regulating system for plate type fuels of Triga-2000 reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Saminto

    2016-01-01

    Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor has been made. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor was made with refer to study result of instrument and control system which is used in BATAN'S reactor. Conceptual design of the control rod regulating system for plate type fuel of TRIGA-2000 reactor consist of 4 segments that is control panel, translator, driver and display. Control panel is used for regulating, safety and display control rod, translator is used for signal processing from control panel, driver is used for driving control rod and display is used for display control rod level position. The translator was designed in 2 modes operation i.e operation by using PLC modules and IC TTL modules. These conceptual design can be used as one of reference of control rod regulating system detail design. (author)

  15. Comparison of performance indicators of different types of reactors based on ISOE database

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2005-01-01

    The optimisation of the operation of a nuclear power plant (NPP) is a challenging issue due to the fact that besides general management issues, a risk associated to nuclear facilities should be included. In order to optimise the radiation protection programmes in around 440 reactors in operation with more than 500 000 monitored workers each year, the international exchange of performance indicators (PI) related to radiation protection issues seems to be essential. Those indicators are a function of a type of a reactor as well as the age and the quality of the management of the reactor. in general three main types of radiation protection PI could be recognised. These are: occupational exposure of workers, public exposure and management of PI related to radioactive waste. The occupational exposure could be efficiently studied using ISOC database. The dependence of occupational exposure on different types of reactors, e.g. PWR, BWR, are given, analysed and compared. (authors)

  16. EPR Flamanville 3, Site Management

    International Nuclear Information System (INIS)

    Menager, Antoine

    2014-01-01

    Antoine Menager, the EPR Flamanville 3 Site Manager described the organization and the management of the Flamanville site during the construction phase. He placed emphasis on Health and Safety, Environmental and Social Responsibility and on Nuclear Safety and Quality

  17. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    International Nuclear Information System (INIS)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won

    2015-01-01

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario

  18. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  19. Revised design for the Tokamak experimental power reactor

    International Nuclear Information System (INIS)

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

    1977-03-01

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

  20. γ radiation level simulation and analysis with MCNP in EPR containment during severe accident

    International Nuclear Information System (INIS)

    Zeng Jun; Liu Shuhuan; Wang Yang; Zhai Liang

    2013-01-01

    The γ dosimetry model based on the EPR core structure, material composition and the designed shielding system was established. The γ-ray dose rate distributions in EPR containment under different conditions including normal operation state, loss-of-coolant accident and core melt severe accident were simulated with MCNP5, and the calculation results under normal operation state and severe accident were compared and analyzed respectively with that of the designed limit. The study results may provide some relative data reference for EPR core accident prediction and reactor accident emergency decision making. (authors)

  1. Experiments utilizing two coupled TRIGA-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Thayer, G [Southern California Edison Co., Rosemead, CA (United States); Jones, B G; Miley, G H [University of Illinois (United States)

    1974-07-01

    An experimental study has been performed on a coupled-core system consisting of two reactors each of which can be made critical by itself, coupled neutronically by a graphite thermal column. Both steady-state and transient measurements were performed on the system. The steady-state measurement consisted of measuring the coupling coefficient between the two reactors. Also, series of measurements were performed while one of the cores was far subcritical and the coupling between the two cores was varied between 1.6 x 10{sup -2} and 1.6 x 10{sup -5} cents by the insertion of a water gap and from 1.6 x 10{sup -2} cents to 6.0 x 10{sup -4} cents by the insertion of a cadmium sheet between the cores. The transient portion of the study was performed by pulsing one of the reactors (the Illinois Advanced TRIGA) and following the pulse into the passive core (the Low Power Reactor Assembly). The first pulse series measured the pulse as it emerged from the thermal column and propagated through the water, where no fuel was present. This provided an analysis of the neutron source to the passive core. The second pulse series was performed with the passive core far subcritical (k{sub eff} {approx_equal} 0.94) and investigated the effects on the transient coupling of the insertion of water gaps of up to 9 inches or a cadmium sheet ({sigma}T = 3.2) between the two cores. Spatial measurements of the pulse in the far subcritical assembly also were performed. The third series of pulses investigated the characteristics of the pulse in the passive core when it was subcritical, just critical, and supercritical, The effects on the FWHM of the pulse in the passive core and on the delay time between the peak of the pulse in the TRIGA and the passive core were measured for the passive core having a k{sub eff} from 0.936 to 1.0015 and the initial period of the pulse in TRIGA varying from 15.6 {+-} .7 ms to 3.58 {+-} .05 ms. The FWHM increased from 13.5 {+-} 0.5 ms to 18.8 {+-} 0.5 ms and delay

  2. Surveillance extension experience at WWER-440 type reactors

    International Nuclear Information System (INIS)

    Gillemot, F.; Uri, G.; Oszwald, F.; Trampus, P.

    1993-01-01

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs

  3. Automatic accounting of nuclear materials at WWER type reactor NPPs

    International Nuclear Information System (INIS)

    Babaev, N.S.; Poznyakov, N.L.; Strelkov, D.F.

    1978-01-01

    The possibilities of automatic accounting of nuclear materials at NPPs based on WWER reactors are considered. Organizational and technical principles of an automated system of accounting that takes into consideration IAEA requirements in conducting accounting documentation are proposed. A program is described for accounting materials using a BESM-6 computer. Operation of the program requires that all accounting data be recorded on conventional carriers of computer information (magnetic tapes, discs, perforated cards), which constitute the basic NPP accounting documents and may be directly used as initial data for a corresponding information program

  4. Surveillance extension experience at WWER-440 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillemot, F; Uri, G [Budapesti Mueszaki Egyetem, Budapest (Hungary); Oszwald, F; Trampus, P

    1994-12-31

    In WWER-440 reactors, the surveillance specimens are located in accelerated irradiation positions. After 5 years, all specimens are withdrawn and the operational changes are not monitored. At Paks NPP a new surveillance program extension has been settled in order to avoid these original program disadvantages and generate further data for plant lifetime management. This paper includes: research performed to prepare the surveillance extension programme, the evaluation method for the surveillance extension, and first results (Charpy and tensile tests). (authors). 6 refs., 12 figs., 3 tabs.

  5. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  6. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  7. Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

    1978-05-01

    Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

  8. Monitoring of core barrel vibrations in WWER type reactor using out-of-reactor ionization chambers

    International Nuclear Information System (INIS)

    Dach, K.

    1982-01-01

    Vibration of the core barrel is least desirable for safe operation of the PWR reactor. These mechanical vibrations are in correlation with the fluctuations of neutron flux density whose time and frequency analysis serves failure diagnosis. The mathematical model is described of the transfer of mechanical vibrations of the core barrel to neutron noise. Other steps are indicated indispensable for the application of the method of neutron noise analysis for in-service diagnostics of nuclear power plants. (Z.M.)

  9. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    Soma, Naoshi; Akimoto, Seiichi; Yokoyama, Iwao.

    1993-01-01

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  10. Accident analysis for new reactor concepts and VVER type reactor design with advanced fuel. STC with Russia. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Mittag, S.; Rohde, U.; Seidel, A.

    2000-10-01

    In the frame of a project on scientific-technical cooperation funded by BMBF/BMWi, the 3D reactor dynamics code DYN3D developed at Forschungszentrum Rossendorf (FZR), has been transferred to the Institute of Physics and Power Engineering (IPPE) Obninsk in Russia and integrated into the software package of IPPE. DYN3D has been coupled to a thermohydraulic system code used in IPPE making available 3D neutron kinetics within this software package. A new macroscopic cross section library has been created using a modified version of the WIMS/D4 code. This library includes data for modernized fuel design containing burnable absorbers in different concentrations, which is tested in VVER-1000 type reactors. The cross section library has been connected to DYN3D. Calculations were performed to check the library in comparison with other data libraries and codes. The code DYN3D and the coupled 3D neutron kinetics/thermal hydraulics code system were used to perform analyses of Anticipated Transients Without Scram (ATWS) for the reactor design ABV-67, an integral reactor concept with small power developed under participation of IPPE. The fluid dynamics code DINCOR developed at IPPE was transferred to FZR. It was used in validation calculations on test problems for the short-term core melt behaviour (CORVIS experiments). (orig.) [de

  11. EPR method for the detection of sensitization in stainless steels

    International Nuclear Information System (INIS)

    Clarke, W.L.; Cowan, R.L.

    1980-01-01

    The overall objective of the program was to improve the reliability of reactor system piping by increasing knowledge of failure causing mechanisms and by enhancing the capability for design evaluation and analysis. Toward the attainment of that objective, a technique has been developed to measure the degree of sensitization quantitatively in thermally treated AISI-304, -304L, -316 and 316L stainless steels. The Electrochemical Potentiokinetic Reactivation (EPR) test was developed because of an industrial need for a rapid, nondestructive, quantitative field test which could be used for assessing sensitization in reactor components. The EPR method consists of developing potentiokinetic curves on a polarized sample obtained by controlled potential sweep from the passive to the active region (reactivation) in a specific electrolyte; details of the test technique have been reported

  12. Theoretical study for ICRF sustained LHD type p-11B reactor

    International Nuclear Information System (INIS)

    Watanabe, Tsuguhiro

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p- 11 B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D- 3 He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p- 11 B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D- 3 He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor (γ HH ). It is shown that high average beta plasma confinement, a large confinement factor (γ HH > 3) and the hot ion mode (T i /T e > 1.4) are necessary to achieve the ignition of the D- 3 He helical reactor. Characteristics of ICRF sustained p- 11 B reactor are analyzed in section 4. The nuclear fusion reaction rate is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p- 11 B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  13. Announcement of recommendations of the Reaktor-Sicherheitskommission. As of 24 July 1997. Joint recommendations of RSK and GPR for safety requirements of future nuclear PWR-type power plants. English versions published in the years 1995 through 1997

    International Nuclear Information System (INIS)

    1997-01-01

    The recommendations, most parts given in English, refer to the European Pressurized Water Reactor (EPR) and have been established by the German RSK (reactor safety commission), the corresponding French organization GPR and the German SSK (radiation protection commission). This publication continues earlier joint recommendations by the national bodies, last published by the German BMU (responsible German ministry) on 5 May 1995, in BAnz. page 7452. The safety recommendations establish the basis for further activities in the Franco-German project for development of the EPR, a PWR type reactor of the next generation. (CB) [de

  14. EPRTM Reactor neutron instrumentation

    International Nuclear Information System (INIS)

    Pfeiffer, Maxime; SALA, Stephanie

    2013-06-01

    The core safety during operation is linked, in particular, to the respect of criteria related to the heat generated in fuel rods and to the heat exchange between the rods and the coolant. This local power information is linked to the power distribution in the core. In order to evaluate the core power distribution, the EPR TM reactor relies on several types of neutron detectors: - ionization chambers located outside the vessel and used for protection and monitoring - a fixed in-core instrumentation based on Cobalt Self Powered Neutron Detectors used for protection and monitoring - a mobile reference in-core instrumentation based on Vanadium aero-balls This document provides a description of this instrumentation and its use in core protection, limitation, monitoring and control functions. In particular, a description of the detectors and the principles of their signal generation is supplied as well as the description of the treatments related to these detectors in the EPR TM reactor I and C systems (including periodical calibration). (authors)

  15. Total decay heat estimates in a proto-type fast reactor

    International Nuclear Information System (INIS)

    Sridharan, M.S.

    2003-01-01

    Full text: In this paper, total decay heat values generated in a proto-type fast reactor are estimated. These values are compared with those of certain fast reactors. Simple analytical fits are also obtained for these values which can serve as a handy and convenient tool in engineering design studies. These decay heat values taken as their ratio to the nominal operating power are, in general, applicable to any typical plutonium based fast reactor and are useful inputs to the design of decay-heat removal systems

  16. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  17. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  18. Modelling the WWER-type reactor dynamics using a hybrid computer. Part 1

    International Nuclear Information System (INIS)

    Karpeta, C.

    Results of simulation studies into reactor and steam generator dynamics of a WWER type power plant are presented. Spatial kinetics of the reactor core is described by a nodal approximation to diffusion equations, xenon poisoning equations and heat transfer equations. The simulation of the reactor model dynamics was performed on a hybrid computer. Models of both a horizontal and a vertical steam generator were developed. The dynamics was investigated over a large range of power by computing the transients on a digital computer. (author)

  19. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

    2000-03-01

    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  20. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  1. The implementation of advanced EPR in the US

    International Nuclear Information System (INIS)

    Hagan, C.; Mico, F.

    2007-01-01

    As electricity demand, clean air concerns and energy prices increase in the U. S., interest in the new nuclear reactors is also on the rise. These factors, along with decades of outstanding performance of the nations existing fleet, have contributed to a favourable market of new nuclear power in the U.S.This article presents some reflexions about the new situation of the market and the response that AREVA offers to he North American Electric utilities: The EPR. (Author)

  2. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  3. Installation of remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials

    International Nuclear Information System (INIS)

    Kato, Yoshiaki; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya; Miwa, Yukio

    2008-06-01

    The remote-handling typed EBSD-OIM analyzer for heavy irradiated reactor materials was installed in the JMTR hot laboratory at the first time in the world. The analyzer is used to study on IASCC (irradiation assisted stress corrosion cracking) or IGSCC (inter granular stress corrosion cracking) in reactor materials. This report describes the measurement procedure, the measured results and the operating experiences on the analyzer in the JMTR hot laboratory. (author)

  4. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    Wach, D.; Mulka, B.; Schnuerer, G.

    1997-01-01

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  5. BH2201 type leakage monitoring equipment of reactor fuel elements

    International Nuclear Information System (INIS)

    Ji Changsong; Dai Zhude; Xie Liangnian; Zhang Shulan; Zhang Shuheng

    1999-01-01

    A high-sensitive equipment monitoring leakage of the reactor fuel elements has been developed. The delayed neutrons emitted from fission product-pioneer nucleus are monitored in the 1st circle water. An array of 3 He proportional counter tubes is designed as a neutron detector for delayed neutrons, the detection geometry of which is near to 4π. In order to reduce the influence of interference factors the monitoring of fission product is carried out with 75s delay. The 87 Br delayed neutron pioneer nucleus is chosen as a monitoring object. The neutron detection efficiency of the developed equipment is 6.1%, which is 3 times higher than one of all available advanced equipment of the same function both at home and abroad

  6. Observation method for inside of FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio; Ishitori, Takashi.

    1992-01-01

    The method of the present invention provides such a method that the surface of a metal case of an ultrasonic transducer keep a good intimate contact with liquid sodium always in a normal state in a short period of time while exhibiting satisfactory wettability. That is, an oxygen concentration in liquid sodium is increased and the inside of the reactor is seen through. Liquid sodium in a state of high oxygen concentration has extremely satisfactory wettability with metals. Accordingly, the metal surface of the ultrasonic transducer can be put to a intimate contact with the liquid metal sodium in a normal state. Further, a coating layer made of nickel or gold is disposed on the surface of the ultrasonic transducer. With such a constitution, the wettability with the liquid metal sodium can further be improved. (I.S.)

  7. Hydraulic Profiling of a Parallel Channel Type Reactor Core

    International Nuclear Information System (INIS)

    Seo, Kyong-Won; Hwang, Dae-Hyun; Lee, Chung-Chan

    2006-01-01

    An advanced reactor core which consisted of closed multiple parallel channels was optimized to maximize the thermal margin of the core. The closed multiple parallel channel configurations have different characteristics to the open channels of conventional PWRs. The channels, usually assemblies, are isolated hydraulically from each other and there is no cross flow between channels. The distribution of inlet flow rate between channels is a very important design parameter in the core because distribution of inlet flow is directly proportional to a margin for a certain hydraulic parameter. The thermal hydraulic parameter may be the boiling margin, maximum fuel temperature, and critical heat flux. The inlet flow distribution of the core was optimized for the boiling margins by grouping the inlet orifices by several hydraulic regions. The procedure is called a hydraulic profiling

  8. Current status of the EPR method to detect irradiated food

    International Nuclear Information System (INIS)

    Desrosiers, M.F.

    1996-01-01

    This review gives a brief outline of the principles of the EPR detection method for irradiated foods by food type. For each food type, the scope, limitations and status of the method are given. The extensive reference list aims to include all which define the method, as well as some rarely cited works of historical importance. (author)

  9. R and D status of an integral type small reactor MRX in JAERI

    International Nuclear Information System (INIS)

    Hoshi, Tsutao; Ochiai, Masaaki; Iida, Hiromasa; Yamaji, Akio; Shimazaki, Junya

    1995-01-01

    JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety study of the thermal hydraulic phenomena as well as the feasibility study of the applicability to merchant ships. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the results of feasibility study. (author)

  10. Status of development - An integral type small reactor MRX in JAERI

    International Nuclear Information System (INIS)

    Hoschi, T.; Ochiai, M.; Shimazaki, J.

    1998-01-01

    JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety systems. Feasibility study and the economical evaluation of nuclear merchant ships have also being performed. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the some results of feasibility study. (author)

  11. Special features of embrittlement of welded joints in shells of VVER-type reactors

    International Nuclear Information System (INIS)

    Kasatkin, O.G.

    1999-01-01

    At present, the atomic power engineering of Russia and Ukraine is based on water-water energy reactors of the VVER-440 and VVER-1000 type, with the electric power of 440 and 1000 MW, respectively. The majority of the VVER-440 reactors are installed in Russia, and VVER-1000 reactors operate in Ukraine. The reactors' shell (RS) is produced from cylindrical shells and a dished end welded together by circular joints under a flux. The RS of the VVER-440 reactor is produced from 15Kh3MFA steel, and the VVER-1000 reactors are produced from 15Kh2NMFA steel. The shell of the VVER-1000 reactor has an internal austenite coating. The condition of the RS metal is determined mainly by the critical brittleness temperature T b at which the impact toughness of specimens with a sharp notch reaches 60 J/cm 2 . The energy reactors, working in western countries, are characterised by a service life of 40 years and discussion is being carried out to extend this lifetime to 60 years. The design service life of the domestic reactors varies from 30 (RS VVER-440) to 40 (RS VVER-1000) years. According to investigations, the service life of the shells of these reactors is restricted by the properties of welded joints which are characterised by higher susceptibility to embrittlement than that of the parent metal, especially due to a higher content in the weld of phosphorus (RS VVER-440) or nickel (RS VVER-1000). Therefore, some experts believe that the actual service life of the RS is shorter than the design life. The accurate evaluation of the service life of welded joints in the RS is very important for the safety of service and also in the economic aspects, because the unjustified decrease of the permissible service life and premature shutdown of units of the nuclear power station result in huge losses

  12. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  13. The EPR - a safe and competitive solution for future energy needs

    International Nuclear Information System (INIS)

    Fischer, U.; Bouteille, F.; Leverenz, R.

    1999-01-01

    NPI, Siemens and Framatome, in co-operation with EDF and the major German Utilities, started the development of the European Pressurized Water Reactor (EPR) as an evolutionary approach. After a careful evaluation of the potential of passive safety features, this way was concluded to be superior compared to a 'revolutionary' approach. The vast majority of advanced reactor designs being developed today is of the evolutionary type. The advantages to base an advanced design on the feedback of operation experience of the more than 100 nuclear power plants designed and constructed by Siemens and Framatome are outstanding. This view is shared by the German and French safety authorities which defined their preference for an evolutionary design early at the beginning of their co-operation for the definition of a common safety approach for future nuclear power plants to be built in Germany and France. In their first common set of recommendations, they gave a clear guideline regarding their point of view for requirements posed to the next generation of nuclear power plants: (1) Preference for an evolutionary design in order to derive a maximum benefit from experience; (2) Significant safety improvements by reduction of core meltdown probability and improvement of the confinement function of the containment under accident conditions; (3) Improvement of operating conditions regarding radiation protection, maintenance and human errors. Besides the French-German co-operation of vendors, utilities and authorities, the European utilities co-operate on a much broader basis for the establishment of the European Utilities' Requirements (EUR). During the development of the basic design, the EPR was continuously assessed against these EUR and it was concluded that the EPR complies with these requirements. At the end of the basic design phase at the end of 1997, all information necessary to file a preliminary safety analysis report and a reliable bill of quantities was elaborated. With

  14. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail: inamgung@kings.ac.kr

    2017-04-01

    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  15. EPR's energy conversion system. Alstom's solutions

    International Nuclear Information System (INIS)

    Ledermann, P.

    2009-01-01

    ARABELLE steam turbines have been developed by Alstom to be used as the energy conversion system of light water reactors with high output power like the N4 PWR and the EPR. ARABELLE turbines cumulate 200.000 hours of service with a reliability ratio of 99.97 per cent. This series of slides presents the main features of the turbine including: the use of the simple flux, the very large shape of low pressure blades, the technology of welded rotors. The other main equipment like the alternator, the condenser, the moisture separator-reheaters, the circulating pumps that Alstom integrates in the energy conversion system have benefited with technological improvements that are also presented. (A.C.)

  16. EPR design tools. Integrated data processing tools

    International Nuclear Information System (INIS)

    Kern, R.

    1997-01-01

    In all technical areas, planning and design have been supported by electronic data processing for many years. New data processing tools had to be developed for the European Pressurized Water Reactor (EPR). The work to be performed was split between KWU and Framatome and laid down in the Basic Design contract. The entire plant was reduced to a logical data structure; the circuit diagrams and flowsheets of the systems were drafted, the central data pool was established, the outlines of building structures were defined, the layout of plant components was planned, and the electrical systems were documented. Also building construction engineering was supported by data processing. The tasks laid down in the Basic Design were completed as so-called milestones. Additional data processing tools also based on the central data pool are required for the phases following after the Basic Design phase, i.e Basic Design Optimization; Detailed Design; Management; Construction, and Commissioning. (orig.) [de

  17. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  18. Roof slab cooling device in a FBR type reactor

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1987-01-01

    Purpose: To obtain a roof slab cooling device capable of retaining cooling performance even in a case of electric power supply stop or failure and effective from economical point of view. Constitution: Atmospheric air is introduced into the cooling chamber of a proof slab and spontaneously passed to a exit pipeway connected to a stack thereby cooling the roof slab. Specifically, atmospheric air entered from the inlet pipeway is introduced to the cooling chamber and absorbs heat generate from the inside of the reactor container. Warmed air is sucked from the exit pipeway and then released into the atmosphere passing through the stack. The air cools the roof slab during circulation due to spontaneous passage and keeps the slab at a low temperature. Since the air is passed spontaneously, no power such as for a blower is required at all and, if the electric power supply should be lost, the cooling power can be maintained as it is to provide a high reliability. Further, since no electric power is required for the blowing power, it has high economical merit. (Horiuchi, T.)

  19. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  20. Method of operating control rods for BWR type reactors

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To eliminate the danger such as fuel element failures due to rapid power increase and form a control rod pattern for obtaining a required power level in a relatively short time. Method: Control rods are disposed so as to vertically enter into and retract from the central region of each four fuel assemblies adjacent to each other respectively. Upon operation of the control rods, every other control rods in the lateral and longitudinal directions among the entire control rods that are inserted completely are extracted completely at the lower flow limit of coolants. Then, the control rods completely inserted are divided into groups inserted deeply and groups inserted less deeply. The less deeply inserted groups are extracted just before the excess of thermal limit value successively in the lower flow limit of the coolants and then the deeply inserted groups are extracted successively till a predetermined power level in the same manner. Therefore, the coolant flow to the reactor core is increased and the power level is raised. (Furukawa, Y.)

  1. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    Otero, D.; Bollini, C.; Sangregorio, D.

    1992-01-01

    We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

  2. Fuel assembly for use in BWR type reactor

    International Nuclear Information System (INIS)

    Inaba, Yuzo.

    1988-01-01

    Purpose: To attain the reduction of neutron irradiation amount to control rods by the improvement in the reactor shutdown margin and the improvement of the control rod worth, by enhancing the arrangement of burnable poisons. Constitution: The number of burnable poison-incorporated fuel rods present in the outer two rows along the sides in adjacent with a control rod among the square lattice arrangement in a fuel assembly is decreased to less than 1/4 for that of total burnable poison-incorporated fuel rods, while the remaining burnable posion-incorporated fuel rods are arranged in the region other than above (that is, those regions not nearer to the control rod). Thus, even if a sufficient number of burnable poison to prolong the controlling effect for the reactivity with the burnable contents as the fuel assembly are disposed, only the burnable poison -incorporated fuel rods by the number less than 1/4 for that of the total burnable poison-incorporated fuel rods are present near the control rod of the fuel assembly. Accordingly, the control rod worth at the initial stage of the burning is increased at both high and normal temperatures. (Kawakami, Y.)

  3. Numerical Investigations of the Influencing Factors on a Rotary Regenerator-Type Catalytic Combustion Reactor

    Directory of Open Access Journals (Sweden)

    Zhenkun Sang

    2018-04-01

    Full Text Available Ultra-low calorific value gas (ULCVG not only poses a problem for environmental pollution, but also createsa waste of energy resources if not utilized. A novel reactor, a rotary regenerator-type catalytic combustion reactor (RRCCR, which integrates the functions of a regenerator and combustor into one component, is proposed for the elimination and utilization of ULCVG. Compared to reversal-flow reactor, the operation of the RRCCR is achieved by incremental rotation rather than by valve control, and it has many outstanding characteristics, such as a compact structure, flexible application, and limited energy for circulation. Due to the effects of the variation of the gas flow and concentration on the performance of the reactor, different inlet velocities and concentrations are analyzed by numerical investigations. The results reveal that the two factors have a major impact on the performance of the reactor. The performance of the reactor is more sensitive to the increase of velocity and the decrease of methane concentration. When the inlet concentration (2%vol. is reduced by 50%, to maintain the methane conversion over 90%, the inlet velocity can be reduced by more than three times. Finally, the highly-efficient and stable operating envelope of the reactor is drawn.

  4. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    International Nuclear Information System (INIS)

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

  5. Effect of Core Configurations on Burn-Up Calculations For MTR Type Reactors

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    Three-dimensional burn-up calculations of MTR-type research reactor were performed using different patterns of control rods , to examine their effect on power density and neutron flux distributions throughout the entire core and on the local burn-up distribution. Calculations were performed using the computer codes' package M TR P C system , using the cell calculation transport code WIMS-D4 and the core calculation diffusion code CITVAP. A depletion study was done and the effects on the reactor fuel were studied, then an empirical formula was generated for every fuel element type, to correlate irradiation to burn-up percentage. Keywords: Neutronic Calculations, Burn-Up, MTR-Type Research Reactors, MTR P C Package, Empirical Formula For Fuel Burn-Up.

  6. Integrity of pressurized water electronuclear reactor vessels. The case of French reactors

    International Nuclear Information System (INIS)

    2012-01-01

    This document aims at identifying elements related to design, manufacturing and control during operation of reactor vessels of the French electronuclear fleet, and more precisely as far as vessel ferrule is concerned. It briefly describes the typical design and elements of most of French PWR vessels with respect to the reactor type (900 MWe, 1300 MWe, 1450 MWe, EPR). It recalls some measures regarding design (for embrittlement assessment) and manufacturing processes (forging operations for shell fabrication, coatings). It discusses the different manufacturing defects which have been noticed (under the coatings, due to hydrogen, and intergranular loss of cohesion due to re-heating). It more particularly comments defects noticed on a Belgium power station reactor in Doel, defects due to hydrogen and some other defects noticed in the French reactor fleet. It presents the different types of control which are performed on vessel shells during operation

  7. Specific public debate on the EPR - Position of 'Sauvons le Climat'

    International Nuclear Information System (INIS)

    2008-01-01

    After having evoked the respective performance of Denmark, Sweden and France in terms of CO 2 emission reduction, electricity consumption per inhabitant, and outlined the unavoidable measures of energy saving and reasonable development of renewable energies while keeping on using nuclear energy, this report more particularly addresses various issues related to the EPR within the perspectives of energy production needs. It also outlines the need to stop the current rate of increase of energy consumption, for resource concerns as well as for CO 2 emission concerns, when renewable energies will remain limited. It discusses the objectives of reduction of greenhouse emissions and states that the EPR could be a determining factor to reach these objectives. Moreover, the EPR technology is an industrial asset for France even in front of the emergence of reactors of fourth generation (high temperature gas reactors, fast breeder reactors). It outlines that the EPR design improves nuclear safety and reduces accident probability. Direct, indirect and external costs of the EPR are discussed. The role of this reactor as a tool for electric power production is finally discussed within the French context and within the European context

  8. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    Kozma, R.; Turkcan, E.; Verhoef, J.P.

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  9. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  10. Activity of corrosion products in pool type reactors with ascending flow in the core

    International Nuclear Information System (INIS)

    Andrade e Silva, Graciete S. de; Queiroz Bogado Leite, Sergio de

    1995-01-01

    A model for the activity of corrosion products in the water of a pool type reactor with ascending flow is presented. The problem is described by a set of coupled differential equations relating the radioisotope concentrations in the core and pool circuits and taking into account two types of radioactive sources: i) those from radioactive species formed in the fuel cladding, control elements, reflector, etc, and afterwards released to the primary stream by corrosion (named reactor sources) and ii) those formed from non radioactive isotopes entering the primary stream by corrosion of the circuit components and being activated when passing through the core (named circuit sources). (author). 6 refs, 3 figs, 4 tabs

  11. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2006-12-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (author)

  12. Neutronic design of a 22 MW MTR type nuclear research reactor

    International Nuclear Information System (INIS)

    Khamis, I.; Khattab, K.; Soleman, I.; Ghazi, N.

    2008-01-01

    The neutronic design calculations of a 22 MW MTR type nuclear research reactor are conducted in this project. This reactor type is selected by the Arab Atomic Energy Commission in a cooperated project. The design calculations are conducted in two methods: The deterministic method, solving the neutron transport and diffusion equations using the WIMSD4 and the CITATION codes, and the probabilistic method using the MCNP code. Good agreements are noticed between the results of the multiplication factor and the neutron flux distribution which prove the accuracy of our models using the two methods. (authors)

  13. Design of the EPR containment

    International Nuclear Information System (INIS)

    Appell, B.; Zaiss, W.

    1996-01-01

    In order to respect the safety objectives set for the EPR (European Pressurized water Reactor), the confinement function must be designed to guard against in the design uncontrolled releases in the environment in all conditions taken into account and to preserve its structural integrity. The concept chosen is a double-wall confinement with technology identical derived from the current French N4 containments and the associated systems ensuring the isolation and the control of leaks. The basic principles aiming to minimize leaks are as follows: no direct leak; state-of-the-art leak tight design of the systems and components passing through the containment building; recovery of potential leaks through the inner wall and the penetration sleeves in the inter-wall space; recovery in the peripheral buildings; and specific measures if necessary. The inner wall is a prestressed concrete shell (55 tendons cables arranged in two horizontal layers and a vertical layer for the barrel) without liner, of free volume 90000 m 3 , an internal diameter of 48 m and 1.3 m thickness. The free volume is chosen so as to rule out the risk of global detonation and the use of catalytic recombiners limits the risk of hydrogen explosion. The design pressure (6.5 bar abs) and temperature of the inner wall are defined for a given volume by the set of three conditions: Pee-4 conditions such as LOCA or SLB, global deflagration of hydrogen and core melt scenario. The rate of leakage in accident conditions from the inner wall must not be higher than 1% per day. The chosen concept must enable satisfactory leak tightness to be preserved for beyond design conditions in order to have margins and to guard against phenomenological uncertainties. The possibility of adding an internal composite liner is being studied. A large scale mockup is being built to validate the hypothesis and methods of leak rates of the inner wall and for the performance testing of the composite liner. The outer wall, made of

  14. Applications of EPR steering in quantum teleportation and NOON states

    Science.gov (United States)

    Zárate, Laura Rosales

    2018-04-01

    Einstein-Podolsky-Rosen (EPR) steering refers to the type of correlations described in the EPR paradox, where one observer seems to affect ("steer") the state of other observer by using local measurements. There have been several works regarding characterization and quantification of EPR steering. One characteristic of this non-local correlation is that it can be asymmetric, while entanglement is symmetric. This asymmetric property is relevant for potential applications of EPR steering to quantum information, in particular to quantum cryptography and quantum teleportation. This latter refers to the process where one observer sends an unknown quantum state to Bob, who is in a different location. They communicate by classical means. Here we will show that EPR steering is a necessary resource to obtain secure continuous variable teleportation. We will also consider NOON states, which is an example of an entangled state. For this state, we will present a steering signature. This contribution reviews the work derived in Refs. [1] and [2], which was presented as an invited talk in ELAF 2017.

  15. Chirp echo Fourier transform EPR-detected NMR.

    Science.gov (United States)

    Wili, Nino; Jeschke, Gunnar

    2018-04-01

    A new ultra-wide band (UWB) pulse EPR method is introduced for observing all nuclear frequencies of a paramagnetic center in a single shot. It is based on burning spectral holes with a high turning angle (HTA) pulse that excites forbidden transitions and subsequent detection of the hole pattern by a chirp echo. We term this method Chirp Echo Epr SpectroscopY (CHEESY)-detected NMR. The approach is a revival of FT EPR-detected NMR. It yields similar spectra and the same type of information as electron-electron double resonance (ELDOR)-detected NMR, but with a multiplex advantage. We apply CHEESY-detected NMR in Q band to nitroxides and correlate the hyperfine spectrum to the EPR spectrum by varying the frequency of the HTA pulse. Furthermore, a selective π pulse before the HTA pulse allows for detecting hyperfine sublevel correlations between transitions of one nucleus and for elucidating the coupling regime, the same information as revealed by the HYSCORE experiment. This is demonstrated on hexaaquamanganese(II). We expect that CHEESY-detected NMR is generally applicable to disordered systems and that our results further motivate the development of EPR spectrometers capable of coherent UWB excitation and detection, especially at higher fields and frequencies. Copyright © 2018 The Author(s). Published by Elsevier Inc. All rights reserved.

  16. Chirp echo Fourier transform EPR-detected NMR

    Science.gov (United States)

    Wili, Nino; Jeschke, Gunnar

    2018-04-01

    A new ultra-wide band (UWB) pulse EPR method is introduced for observing all nuclear frequencies of a paramagnetic center in a single shot. It is based on burning spectral holes with a high turning angle (HTA) pulse that excites forbidden transitions and subsequent detection of the hole pattern by a chirp echo. We term this method Chirp Echo Epr SpectroscopY (CHEESY)-detected NMR. The approach is a revival of FT EPR-detected NMR. It yields similar spectra and the same type of information as electron-electron double resonance (ELDOR)-detected NMR, but with a multiplex advantage. We apply CHEESY-detected NMR in Q band to nitroxides and correlate the hyperfine spectrum to the EPR spectrum by varying the frequency of the HTA pulse. Furthermore, a selective π pulse before the HTA pulse allows for detecting hyperfine sublevel correlations between transitions of one nucleus and for elucidating the coupling regime, the same information as revealed by the HYSCORE experiment. This is demonstrated on hexaaquamanganese(II). We expect that CHEESY-detected NMR is generally applicable to disordered systems and that our results further motivate the development of EPR spectrometers capable of coherent UWB excitation and detection, especially at higher fields and frequencies.

  17. Development of a new dosimeter of EPR based on lactose

    International Nuclear Information System (INIS)

    Cruz C, L.; Torijano C, E.; Azorin N, J.; Aguirre G, F.; Cruz Z, E.

    2014-08-01

    50 years have passed since was proposed using the amino acid alanine as dosimeter advantage the phenomenon of electron paramagnetic resonance (EPR); this dosimetric method has reached a highly competitive level regarding others dosimetry classic methods, for example the thermoluminescence or the use of Fricke dosimeters, to measure high dose of radiation. In this type of materials, the free radicals induced by the radiation are stable and their concentration is proportional to the absorbed dose may be determined by the amplitude pick to pick of the first derived of the EPR absorption spectrum. The obtained results studying the EPR response of lactose tablets elaborated in the Universidad Autonoma Metropolitana, Unidad Iztapalapa are presented. The tablets were irradiated with gamma radiation of 60 Co in the irradiator Gamma beam 651-Pt of the Instituto de Ciencias Nucleares de la Universidad Nacional Autonoma de Mexico to a dose rate of 8 kGy-h -1 and their EPR response in a EPR spectrometer e-scan Bruker. The obtained response in function of the dose was lineal in the interval of 1 at 10 kGy. The lactose sensibility was compared with the l-alanine, used as reference, and the result was consistently 0.25 of this. Due to the linearity shown in the interval of used dose and their low production cost, we conclude that the lactose is a promissory option for the dosimetry of high dose of radiation. (author)

  18. Rapid-scan EPR imaging.

    Science.gov (United States)

    Eaton, Sandra S; Shi, Yilin; Woodcock, Lukas; Buchanan, Laura A; McPeak, Joseph; Quine, Richard W; Rinard, George A; Epel, Boris; Halpern, Howard J; Eaton, Gareth R

    2017-07-01

    In rapid-scan EPR the magnetic field or frequency is repeatedly scanned through the spectrum at rates that are much faster than in conventional continuous wave EPR. The signal is directly-detected with a mixer at the source frequency. Rapid-scan EPR is particularly advantageous when the scan rate through resonance is fast relative to electron spin relaxation rates. In such scans, there may be oscillations on the trailing edge of the spectrum. These oscillations can be removed by mathematical deconvolution to recover the slow-scan absorption spectrum. In cases of inhomogeneous broadening, the oscillations may interfere destructively to the extent that they are not visible. The deconvolution can be used even when it is not required, so spectra can be obtained in which some portions of the spectrum are in the rapid-scan regime and some are not. The technology developed for rapid-scan EPR can be applied generally so long as spectra are obtained in the linear response region. The detection of the full spectrum in each scan, the ability to use higher microwave power without saturation, and the noise filtering inherent in coherent averaging results in substantial improvement in signal-to-noise relative to conventional continuous wave spectroscopy, which is particularly advantageous for low-frequency EPR imaging. This overview describes the principles of rapid-scan EPR and the hardware used to generate the spectra. Examples are provided of its application to imaging of nitroxide radicals, diradicals, and spin-trapped radicals at a Larmor frequency of ca. 250MHz. Copyright © 2017 Elsevier Inc. All rights reserved.

  19. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  20. Manning designs for nuclear district-heating plant (NDHP) with RUTA-type reactor

    International Nuclear Information System (INIS)

    Gerasimova, V.S.; Mikhan, V.I.; Romenkov, A.A.

    2001-01-01

    RUTA-type reactor is a water cooled water-moderated pool-type reactor with an atmospheric pressure air medium. The reactor has been designed for heating and hot water supply. Nuclear district heating plant (NDHP) with RUTA-type reactor facility has been designed with a three circuit layout. Primary circuit components are arranged integrally in the reactor vessel. Natural coolant circulation mode is used in the primary circuit. A peculiarity of RUTA-based NDHP as engineered system is a smooth nature of its running slow variation of the parameters at transients. Necessary automation with application of computer equipment will be provided for control and monitoring of heat production process at NDHP. Under developing RUTA-based NDHP it is foreseen that operating staff performs control and monitoring of heat generation process and heat output to consumers as well as current maintenance of NDHP components. All other works associated with NDHP operation should be fulfilled by extraneous personnel. In so doing the participation of operating staff is also possible. (author)

  1. EPR: Some History and Clarification

    Science.gov (United States)

    Fine, Arthur

    2002-04-01

    Locality, separation and entanglement 1930s style. We’ll explore the background to the 1935 paper by Einstein, Podolsky and Rosen, how it was composed, the actual argument of the paper, the principles used, and how the paper was received by Schroedinger, and others.We’ll also look at Bohr’s response: the extent to which Bohr connects with what Einstein was after in EPR and the extent to EPR marks a shift in Bohr’s thinking about the quantum theory.

  2. EPR in B physics and elsewhere

    International Nuclear Information System (INIS)

    Lipkin, H.J.; Tel Aviv Univ.; Argonne National Lab., IL

    1997-01-01

    The application of Einstein-Podolsky-Rosen correlations in Υ(4s) → B anti B decays to research in CP violation is the first and probably only use of EPR as a technique for research in new physics. Elsewhere highly sophisticated EPR projects question EPR and test its predictions to look for violations of quantum mechanics, hidden variables, Bell''s inequalities, etc

  3. A nuclear desalination complex with a VK-300 boiling type reactor facility

    International Nuclear Information System (INIS)

    Kuznetzov, Y.N.; Mishanina, Y.A.; Romenkov, A.A.

    2004-01-01

    RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m 3 /day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

  4. Wave-guide type photo reactor for water purification

    International Nuclear Information System (INIS)

    Nobuaki, Negishi; Feng, He; Sadao, Matsuzawa; Koji, Takeuchi; Kayo, Ohno

    2006-01-01

    A wave-guide type photo-catalytic rod that is consisting of a glass tube with transparent TiO 2 (outside) and an optical wave-guide rod (inside) was designed and examined its performance. A model of polluted water, which contains 100 ppm of toluene or phenol, was taken in a 500 ml of beaker and the performance of this unit was evaluated by the removal rate of pollutants in water under photo-irradiation. Acrylic rod with 6-mm diameter was used as the wave-guide of light. One end of acrylic rod 50 mm had a frosted part or a screw thread for increasing seep out of the light. For the glass tube with transparent TiO 2 , four kinds with different film thickness were prepared by the dip-coating method. The wave-guide type photo-catalytic rods effectively eliminated toluene and phenol and the total amount of intermediates formation was low. (authors)

  5. EPR: the nuclear impasse; EPR: l'impasse nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Marillier, F [Association Ecologiste Greenpeace (France)

    2008-07-01

    The questions relative to the climatic change constitute crucial challenges for the next ten years. In this context the author aims to show how the EPR project illustrates the nuclear french ''autism''. He presents and analyzes the international and environmental impacts of this obsolete technology, as a project useless and dangerous. (A.L.B.)

  6. EPR: the nuclear impasse; EPR: l'impasse nucleaire

    Energy Technology Data Exchange (ETDEWEB)

    Marillier, F. [Association Ecologiste Greenpeace (France)

    2008-07-01

    The questions relative to the climatic change constitute crucial challenges for the next ten years. In this context the author aims to show how the EPR project illustrates the nuclear french ''autism''. He presents and analyzes the international and environmental impacts of this obsolete technology, as a project useless and dangerous. (A.L.B.)

  7. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K.; Kim, J. H.

    2015-01-01

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed

  8. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  9. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  10. A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

    International Nuclear Information System (INIS)

    Ardaneh, Kazem; Zaferanlouei, Salman

    2013-01-01

    Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

  11. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro, E-mail: duvan.castellanos@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: pedro.rossi@ufabc.edu.br, E-mail: pedro.carajilescov10@gmail.com [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil). Centro de Engenharias, Modelagem e Ciências Sociais Aplicadas

    2017-07-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  12. Thermal-hydraulic code for estimating safety limits of nuclear reactors with plate type fuels

    International Nuclear Information System (INIS)

    Castellanos, Duvan A.; Moreira, João L.; Maiorino, Jose R.; Rossi, Pedro R.; Carajilescov, Pedro

    2017-01-01

    To ensure the normal and safe operation of PWR type nuclear reactors is necessary the knowledge of nuclear and heat transfer properties of the fuel, coolant and structural materials. The thermal-hydraulic analysis of nuclear reactors yields parameters such as the distribution of fuel and coolant temperatures, and the departure from nucleated boiling ratio. Usually computational codes are used to analyze the safety performance of the core. This research work presents a computer code for performing thermal-hydraulic analyses of nuclear reactors with plate-type fuel elements operating at low pressure and temperature (research reactors) or high temperature and pressure (naval propulsion or small power reactors). The code uses the sub-channel method based on geometric and thermal-hydraulic conditions. In order to solve the conservation equations for mass, momentum and energy, each sub-channel is divided into control volumes in the axial direction. The mass flow distribution for each fuel element of core is obtained. Analysis of critical heat flux is performed in the hottest channel. The code considers the radial symmetry and the chain or cascade method for two steps in order to facilitate the whole analysis. In the first step, we divide the core into channels with size equivalent to a fuel assembly. >From this analysis, the channel with the largest enthalpy is identified as the hot assembly. In the second step, we divide the hottest fuel assembly into sub-channels with size equivalent to one actual coolant channel. As in the previous step, the sub-channel with largest final enthalpy is identified as the hottest sub-channel. For the code validation, we considered results from the chinese CARR research reactor. The code reproduced well the CARR reactor results, yielding detailed information such as static pressure in the channel, mass flow rate distribution among the fuel channels, coolant, clad and centerline fuel temperatures, quality and local heat and critical heat

  13. Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Tsuguhiro (ed.)

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  14. Controlled thermonuclear fusion in TOKAMAK type reactors, the European example: Joint European Torus (JET)

    International Nuclear Information System (INIS)

    Paris, P.J.; Yassen, F.; Assis, A.S. de; Raposo, C.

    1988-07-01

    The development of controlled thermonuclear reaction in TOKAMAK type reactors, and the main projects in the world are presented. The main characteristics of the JET (Joint European Torus) program, the perspectives for energy production, and the international cooperation for viable use of the TOKAMAK are analysed. (M.C.K.) [pt

  15. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  16. Programme of hot points eradication (Co-60) led on French PWR type reactors

    International Nuclear Information System (INIS)

    Rocher, A.; Ridoux, P.; Anthoni, S.; Brun, C.

    1998-01-01

    The question of hot points (pellets rich in cobalt 59 or in cobalt 60 in a PWR type reactor), is studied from the radiation protection point of view. The purpose is to see how to optimize the radiation protection, the elimination of these hot points can bring an improvement. (N.C.)

  17. Roof loading and response following a HCDA in a pool-type reactor

    International Nuclear Information System (INIS)

    Lancefield, M.J.; Leigh, K.M.; Potter, R.; Staniforth, R.

    1979-01-01

    In a pool-type reactor the loading and response of the roof structure to a HCDA is important to safety analysis and design. The U.K. programme of experimental and theoretical work on this topic is described. Good progress in understanding and evaluating the complex processes has been made and this is illustrated by results from experimental and theoretical work. 5 refs

  18. Modernization for safety purposes of Russian nuclear power plants with channel-type reactors

    International Nuclear Information System (INIS)

    Riakhin, V.M.

    1999-01-01

    The nineties have crucially changed the Russian policy towards channel-type reactors known as RBMK. After the period of intensive commissioning the new Units (Kursk NPP: 1976, 1979, 1983,1985; Smolensk NPP 1982, 1985, 1990), the main financial flow was directed into reconstruction of these units. Safety upgrade of the units of Kursk NPP is presented in more details

  19. Control console conceptual design for sheet type fuels of Triga Mark-II reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Kurnia Wibowo; Anang Susanto

    2016-01-01

    The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

  20. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.