WorldWideScience

Sample records for assisted reactor concept

  1. Fission fragment assisted reactor concept for space propulsion: Foil reactor

    International Nuclear Information System (INIS)

    Wright, S.A.

    1991-01-01

    The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures

  2. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

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

  3. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  4. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

    Molten salt reactor is an advanced breeder concept which is suited for the utilization of thorium for nuclear power production. This reactor is based on the use of solutions of uranium or plutonium fluorides in LiF-BeF 2 -ThF 4 as fuel. Unlike the conventional reactors, no external coolant is used in the reactor core and the fuel salt itself is circulated through heat exchangers to transfer the fission produced heat to a secondary salt (NaF-NaBF 4 ) for steam generation. A part of the fuel stream is continuously processed to isolate 233 Pa, so that it can decay to fissile 233 U without getting converted to 234 Pa, and for the removal of neutron absorbing fission products. This on-line processing scheme makes this reactor concept to achieve a breeding ratio of 1.07 which is the highest for any thermal breeder reactor. Experimental studies at the Bhabha Atomic Research Centre, Bombay, have established the use of plutonium as fuel for this reactor. This molten salt reactor concept is described and the work conducted at the Bhabha Atomic Research Centre is summarised. (auth.)

  5. Integral Fast Reactor concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path

  6. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  7. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Furtek, A.

    2008-01-01

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

  8. The integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Yoon I.; Marchaterre, J.F.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) an integral fuel cycle, based on pyrometallurgical processing and injection-cast fuel fabrication, with the fuel cycle facility collocated with the reactor, if so desired. This paper gives a review of the IFR concept

  9. New fast reactor installation concept

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    The large size and complexity of fast reactor installations are emphasised and these difficulties will be increased with the advent of fast reactors of higher power. In this connection a new concept of fast reactor installation is described with a view to reducing the size of the installation and enabling most components, including even the primary vessel, to be constructed within the confines of a workshop. Full constructional details are given. (U.K.)

  10. Reactor concepts for laser fusion

    International Nuclear Information System (INIS)

    Meier, W.R.; Maniscalco, J.A.

    1977-07-01

    Scoping studies were initiated to identify attractive reactor concepts for producing electric power with laser fusion. Several exploratory reactor concepts were developed and are being subjected to our criteria for comparing long-range sources of electrical energy: abundance, social costs, technical feasibility, and economic competitiveness. The exploratory concepts include: a liquid-lithium-cooled stainless steel manifold, a gas-cooled graphite manifold, and fluidized wall concepts, such as a liquid lithium ''waterfall'', and a ceramic-lithium pellet ''waterfall''. Two of the major reactor vessel problems affecting the technical feasibility of a laser fusion power plant are: the effects of high-energy neutrons and cyclical stresses on the blanket structure and the effects of x-rays and debris from the fusion microexplosion on the first-wall. The liquid lithium ''waterfall'' concept is presented here in more detail as an approach which effectively deals with these damaging effects

  11. The Integral Fast Reactor concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes the key features and potential advantages of the IFR concept, its technology development status, fuel cycle economics potential, and its future development path

  12. Integral fast reactor concept

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Sevy, R.H.

    1984-01-01

    Key features of the IFR consist of a pool-type plant arrangement, a metal fuel-based core design, and an integral fuel cycle with colocated fuel cycle facility. Both the basic concept and the technology base have been demonstrated through actual integral cycle operation in EBR-II. This paper discusses the inherent safety characteristics of the IFR concept

  13. Fast mixed spectrum reactor concept

    International Nuclear Information System (INIS)

    Kouts, H.J.C.; Fischer, G.J.; Cerbone, R.J.

    1979-04-01

    The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high

  14. The advanced MAPLE reactor concept

    International Nuclear Information System (INIS)

    Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

    1989-01-01

    High-flux neutron sources are continuing to be of interest both in Canada and internationally to support materials testing for advanced power reactors, new developments in extracted-neutron-beam applications, and commercial production of selected radioisotopes. The advanced MAPLE reactor concept has been developed to meet these needs. The advanced MAPLE reactor is a new tank-type D 2 O reactor that uses rodded low-enrichment uranium fuel in a compact annular core to generate peak thermal-neutron fluxes of 1 x 10 19 n·s -1 in a central irradiation rig with a thermal power output of 50 MW. Capital and incremental development costs are minimized by using MAPLE reactor technology to the greatest extent practicable

  15. The advanced MAPLE reactor concept

    International Nuclear Information System (INIS)

    Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

    1989-01-01

    In Canada the need for advanced neutron sources has long been recognized. During the past several years Atomic Energy of Canada Limited (AECL) has been developing the new MAPLE multipurpose reactor concept. To date, the MAPLE program has focused on the development of a modest-cost multipurpose medium-flux neutron source to meet contemporary requirements for applied and basic research using neutron beams, for small-scale materials testing and analysis and for radioisotope production. The basic MAPLE concept incorporates a compact light-water cooled and moderated core within a heavy water primary reflector to generate strong neutron flux levels in a variety of irradiation facilities. In view of renewed Canadian interest in a high-flux neutron source, the MAPLE group has begun to explore advanced concepts based on AECL's experience with heavy water reactors. The overall objective is to define a high-flux facility that will support materials testing for advanced power reactors, new developments in extracted neutron-beam applications, and/or production of radioisotopes. The design target is to attain performance levels of HFR-Grenoble, HFBR, HFIR in a new heavy water-cooled, -moderated,-reflected reactor based on rodded LEU fuel. Physics, shielding, and thermohydraulic studies have been performed for the MAPLE heavy water reactor. 14 refs., 4 figs., 1 tab

  16. The advanced MAPLE reactor concept

    International Nuclear Information System (INIS)

    Lidstone, R.F.; Lee, A.G.; Gillespie, G.E.; Smith, H.J.

    1989-01-01

    During the past several years, Atomic Energy of Canada Limited (AECL) has been developing the new MAPLE multipurpose reactor concept, which is capable of generating peak thermal neutron fluxes of up to 3 x 10 18 n/m 2 s in its heavy water reflector at a nominal thermal power level of 15MW. An assessment of the MAPLE-D 2 O reactor has shown that it could also be used as a high-flux neutron source. it could be developed to be used for several applications if a 12-site annular core is used. Thermal fluxes several times greater than in existing facilities would be available (author)

  17. Electron beam solenoid reactor concept

    International Nuclear Information System (INIS)

    Bailey, V.; Benford, J.; Cooper, R.; Dakin, D.; Ecker, B.; Lopez, O.; Putman, S.; Young, T.S.T.

    1977-01-01

    The electron Beam Heated Solenoid (EBHS) reactor is a linear magnetically confined fusion device in which the bulk or all of the heating is provided by a relativistic electron beam (REB). The high efficiency and established technology of the REB generator and the ability to vary the coupling length make this heating technique compatible with several radial and axial enery loss reduction options including multiple-mirrors, electrostatic and gas end-plug techniques. This paper addresses several of the fundamental technical issues and provides a current evaluation of the concept. The enhanced confinement of the high energy plasma ions due to nonadiabatic scattering in the multiple mirror geometry indicates the possibility of reactors of the 150 to 300 meter length operating at temperatures > 10 keV. A 275 meter EBHS reactor with a plasma Q of 11.3 requiring 33 MJ of beam eneergy is presented

  18. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1988-06-01

    The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

  19. Modular Stellarator Fusion Reactor concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-08-01

    A preliminary conceptual study is made of the Modular Stellarator Reactor (MSR). A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. The physics basis of the design point is described together with supporting magnetics, coil-force, and stress computations. The approach and results presented herein will be modified in the course of ongoing work to form a firmer basis for a detailed conceptual design of the MSR

  20. Iser: an international inherently safe reactor concept

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki

    1988-01-01

    Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

  1. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  2. A nuclear power reactor concept for Brazil

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1980-01-01

    For the purpose of developing an independent national nuclear technology and effective manner of transferring such a technology, as well as developing a modern reactor, a new nuclear power reactor concept is proposed which is considered as a suitable and viable project for Brazil to support its development and finally construct its prototype as an indigeneous venture. (Author) [pt

  3. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  4. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  5. Development of physical conceptions of fast reactors

    International Nuclear Information System (INIS)

    Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

    2013-01-01

    • Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

  6. A new advanced safe nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    1999-01-01

    The reactor design is based on fluidized bed concept and utilizes pressurized water reactor technology. The fuel is automatically removed from the reactor by gravity under any accident condition. The reactor demonstrates the characteristics of inherent safety and passive cooling. Here two options for modification to the original design are proposed in order to increase the stability and thermal efficiency of the reactor. A modified version of the reactor involves the choice of supercritical steam as the coolant to produce a plant thermal efficiency of about 40%. Another is to modify the shape of the reactor core to produce a non-fluctuating bed and consequently guarantee the dynamic stability of the reactor. The mixing of Tantalum in the fuel is also proposed as an additional inhibition to power excursion. The spent fuel pellets may not be considered nuclear waste since they are in the shape and size that can easily be used as a a radioactive source for food irradiation and industrial applications. The reactor can easily operate with any desired spectrum by varying the porosity in order to be a plutonium burner or utilize a thorium fuel cycle. (author)

  7. SEBREZ: an inertial-fusion-reactor concept

    International Nuclear Information System (INIS)

    Meier, W.R.

    1982-01-01

    The neutronic aspects of an inertial fusion reactor concept that relies on asymmetrical neutronic effects to enhance the tritium production in the breeding zones have been studied. We find that it is possible to obtain a tritium breeding ratio greater than 1.0 with a chamber configuration in which the breeding zones subtend only a fraction of the total solid angle. This is the origin of the name SEBREZ which stands for SEgregated BREeding Zones. It should be emphasized that this is not a reactor design study; rather this study illustrates certain neutronic effects in the context of a particular reactor concept. An understanding of these effects forms the basis of a design technique which has broader application than just the SEBREZ concept

  8. Nuclear Burning Wave Modular Fast Reactor Concept

    International Nuclear Information System (INIS)

    Kodochigov, N.G.; Sukharev, Yu.P.

    2014-01-01

    The necessity to provide nuclear power industry, comparable in a scope with power industry based on a traditional fuel, inspired studies of an open-cycle fast reactor aimed at: - solution of the problem of fuel provision by implementing the highest breeding characteristics of new fissile materials of raw isotopes in a fast reactor and applying accumulated fissile isotopes in the same reactor, independently on a spent fuel reprocessing rate in the external fuel cycle; - application of natural or depleted uranium for makeup fuel, which, with no spent fuel reprocessing, forms the most favorable non-proliferation conditions; - application of inherent properties of the core and reactor for safety provision. The present report, based on previously published papers, gives the theoretical backgrounds of the concept of the reactor with a nuclear burning wave, in which an enriched-fuel core (driver) is replaced by a blanket, and basic conditions for nuclear burning wave initiating and keeping are shown. (author)

  9. Compact magnetic fusin reactor concepts

    International Nuclear Information System (INIS)

    Chung, K.M.

    1984-01-01

    Compact, high-power-density approaches to fusion power represent alternatives to main-line fusion concepts, Tokamaks and mirrors. If technological issues are resolved, theses approaches would yield small, low-cost fusion power plants. This survey reviews the principal physics and technology employed by leading compact magnetic fusion plants. (Author)

  10. On blanket concepts of the Helias reactor

    International Nuclear Information System (INIS)

    Wobig, H.; Harmeyer, E.; Herrnegger, F.; Kisslinger, J.

    1999-07-01

    The paper discusses various options for a blanket of the Helias reactor HSR22. The Helias reactor is an upgrade version of the Wendelstein 7-X device. The dimensions of the Helias reactor are: major radius 22 m, average plasma radius 1.8 m, magnetic field on axis 4.75 T, maximum field 10 T, number of field periods 5, fusion power 3000 MW. The minimum distance between plasma and coils is 1.5 m, leaving sufficient space for a blanket and shield. Three options of a breeding blanket are discussed taking into account the specific properties of the Helias configuration. Due to the large area of the first wall (2600 m 2 ) the average neutron power load on the first wall is below 1 MWm .2 , which has a strong impact on the blanket performance with respect to lifetime and cooling requirements. A comparison with a tokamak reactor shows that the lifetime of first wall components and blanket components in the Helias reactor is expected to be at least two times longer. The blanket concepts being discussed in the following are: the solid breeder concept (HCPB), the dual-coolant Pb-17Li blanket concept and the water-cooled Pb-17Li concept (WCLL). (orig.)

  11. Light ion driven inertial fusion reactor concepts

    International Nuclear Information System (INIS)

    Cook, D.L.; Sweeney, M.A.; Buttram, M.T.; Prestwich, K.R.; Moses, G.A.; peterson, R.R.; Lovell, E.G.; Englestad, R.L.

    1980-01-01

    The possibility of designing fusion reactor systems using intense beams of light ions has been investigated. concepts for beam production, transport, and focusing on target have been analyzed in light of more conservative target performance estimates. Analyses of the major criteria which govern the design of the beam-target-cavity tried indicate the feasibility of designing power systems at the few hundred megawatt (electric) level. This paper discusses light ion fusion reactor (LIFR) concepts and presents an assessment of the design limitations through quantitative examples

  12. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

    1987-01-01

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  13. New reactors concepts and scenarios

    International Nuclear Information System (INIS)

    Gandini, A.

    2001-01-01

    In recent years an increasing interest is observed with respect to subcritical, accelerator driven systems (ADS), for their possible role in perspective future nuclear energy scenarios, as actinide (Pu and MA) incinerators, and/or claimed energy plants with potential enhanced safety characteristics. Important research programs are devoted to the various related fields of research. Extensive studies on the ADS behavior under incidental conditions are in particular made, for verifying their claimed advantage, under the safety point of view, with respect to the corresponding critical reactors. Corresponding medium and long range scenarios are being studied to cope with a number of concerns associated with the safety (power excursions. residual heat risk), as well as with the fuel flow (criticality accidents, fuel diversion, radiological risk, proliferation). In the present work we shall try to review current lines of research in this field, and comment on possible scenarios so far envisaged. (author)

  14. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

    1992-05-01

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  15. Fetal habituation in assisted conception.

    Science.gov (United States)

    Joy, Jolly; McClure, Neil; Hepper, Peter G; Cooke, Inez

    2012-06-01

    Neurodevelopment outcomes of children conceived by Assisted Reproductive Technology (ART)have been the subject of much recent attention. To date there are no reports of neurodevelopmental performance before birth in this group. To compare habituation (a measure of brain function) in fetuses conceived by assisted reproduction techniques (ART) with naturally conceived (NC) fetuses. Case control study. Women with singleton pregnancies matched for maternal age, parity and smoking were recruited in 2 groups: ART (n=20) and NC (n=20). Sound stimuli (250 Hz, 110 dB) at 10 second intervals lasting 2 s were administered to the fetus. The end point was habituation (cessation of movement for five consecutive stimuli) or a maximum of 30 stimuli. Responses of the fetus were observed with ultrasound at 28, 32 and 36 weeks' gestation, video-recorded and anonymised for analysis. At 28 weeks' gestation significantly more ART fetuses responded to sound of 250 Hz, 110 dB (p=0.02) but this difference did not persist at 32 and 36 weeks'. There was a significant increase in nonresponders as gestation advanced in the ART group. There was no difference in habituation or mean number of trials to habituate at all three gestations. ART fetuses demonstrated no differences in habituation suggesting that there is no neurodevelopment delay. However, a decrease in response to sound as gestation advances might be a harbinger for poor perinatal outcomes and needs exploration. Crown Copyright © 2011. Published by Elsevier Ireland Ltd. All rights reserved.

  16. ULTRA SCWR+: Practical advanced water reactor concepts

    International Nuclear Information System (INIS)

    Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

    2008-01-01

    Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

  17. PIUS principle and the SECURE reactor concepts

    International Nuclear Information System (INIS)

    Hannerz, K.

    1987-01-01

    The author introduces the SECURE reactor concept, a reactor intended for producing heat for district heating grids, desalination, and certain process industries. A detailed design of a 400 MWth plant has been completed and is being offered commercially. The authors present first, a summary of the current situation and then the design philosophy of the SECURE reactor concepts. The authors propose a design based on a light water reactor, as opposed to high temperature gas cooled reactor, but introduce new features which are designed to eliminate the element of human error in preparing for and handling emergencies. The authors propose two rules to avoid overheating, i.e.., the PIUS design principle, which are: to keep the core submerged in water; and to ensure that the rate of heat generation in the submerged core is low enough to avoid overheating of the fuel (dryout). The acronym PIUS stands for Process Inherent Ultimate Safety. A detailed system modeling is given of the PIUS primary system. The design of the plant is divided into two parts: the nuclear island, which is comprised of the concrete vessel and its contents; and the balance of the plant, which is comprised of all other components, including the turbine plant

  18. Modular Stellarator Fusion Reactor (MSR) concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1981-01-01

    A preliminary conceptual study has been made of the Modulator Stellarator Reactor (MSR) as a stedy-state, ignited, DT-fueled, magnetic fusion reactor. The MSR concept combines the physics of classic stellarator confinement with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4.8-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an l = 2 system with a plasma aspect ratio of 11. Neither an economic analysis nor a detailed conceptual engineering design is presented here, as the primary intent of this scoping study is the elucidation of key physics tradeoffs, constraints, and uncertainties for the ultimate power-reactor embodiment

  19. Particle bed reactor nuclear rocket concept

    International Nuclear Information System (INIS)

    Ludewig, H.

    1991-01-01

    The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

  20. Molten salt reactors. The AMSTER concept

    International Nuclear Information System (INIS)

    Vergnes, J.; Garzenne, C.; Lecarpentier, D.; Mouney, H.

    2001-01-01

    This article presents the concept of actinide molten salt transmuter (AMSTER). This reactor is graphite-moderated and is dedicated to the burning of actinides. The main difference with a molten salt reactor is that its liquid fuel undergoes an on-line partial reprocessing in which fission products are extracted and heavy nuclei are reintroduced into the fuel. In order to maintain the reactivity regular injections of 235 U-salt are made. In classical reactors, fuel burn-up is limited by the swelling of the cladding and the radiation fuel pellets resistance, in AMSTER there is no limitation to the irradiation time of the fuel, so all the actinides can be burnt or transmuted. (A.C.)

  1. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  2. Trends and developments in magnetic confinement fusion reactor concepts

    International Nuclear Information System (INIS)

    Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

    1981-01-01

    An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

  3. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  4. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  5. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  6. Containment concepts assessment for the SEAFP reactor

    International Nuclear Information System (INIS)

    Di Pace, L.; Natalizio, A.

    2000-01-01

    A simple methodology has been developed for making relative comparisons of potential containment designs for future fusion reactors. The assessment methodology requires only conceptual design information. The application of this methodology, at the early stages of a fusion reactor design, provides designers useful information regarding the suitability of various containment designs and design features. Because the radiation hazard from the operation of future fusion power reactors is expected to be low, the containment design, in addition to public safety, needs to take into account worker safety considerations, as well as factors important to the reliable and economical operation of the power plant. Several containment concepts have been assessed with a methodology that takes into account public safety, worker safety, operability and maintainability as well as cost. This paper describes this methodology and presents the results of the assessment. The paper concludes that, to obtain a containment design that is optimised with respect to safety, operational and cost factors, designers should focus on a containment that is conceptually simple-that is, one utilising a single, large containment building without relying on special features such as expansion volumes, pressure suppression pools or spray systems

  7. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  8. Assessment of nuclear reactor concepts for low power space applications

    Science.gov (United States)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  9. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    Popp, P.

    1987-01-01

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  10. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  11. New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

    1998-11-01

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

  12. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    International Nuclear Information System (INIS)

    Allen, Francis; Bonin, Hugues

    2008-01-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  13. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

    2008-07-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  14. The concept of the innovative power reactor

    Directory of Open Access Journals (Sweden)

    Sang Won Lee

    2017-10-01

    Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

  15. An energy amplifier fluidized bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, F.; Seifritz, W.

    2001-01-01

    The concept of a fluidized bed nuclear reactor driven by an energy amplifier system is described. The reactor has promising characteristics of inherent safety and passive cooling. The reactor can easily operate with any desired spectrum in order to be a plutonium burner or have it operate with thorium fuel cycle. (orig.) [de

  16. Review of the current status of linear hybrid reactor concepts

    International Nuclear Information System (INIS)

    Schultz, K.R.

    1977-07-01

    A review was made of the current status of linear fusion-fission hybrid reactor design studies in the USA. The linear hybrid reactor concepts reviewed include the linear theta-pinch hybrid reactor being studied at Los Alamos Scientific Laboratory, the electron beam-heated solenoid hybrid reactor under development at Physics International Co., the laser-heated solenoid hybrid reactor being investigated at Mathematical Sciences Northwest, Inc., and the linear fusion waste burning reactor being studied at General Atomic Company. The discussion addresses confinement and heating mechanisms for each concept, as well as the hybrid blanket designs. The current state of the four reactor designs is summarized and the performance of the various concepts compared

  17. Preliminary design concepts of an advanced integral reactor

    International Nuclear Information System (INIS)

    Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

    1997-01-01

    An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

  18. A computer code for Tokamak reactor concepts evaluation

    International Nuclear Information System (INIS)

    Rosatelli, F.; Raia, G.

    1985-01-01

    A computer package has been developed which could preliminarily investigate the engineering configuration of a tokamak reactor concept. The code is essentially intended to synthesize, starting from a set of geometrical and plasma physics parameters and the required performances and objectives, three fundamental components of a tokamak reactor core: blanket+shield, TF magnet, PF magnet. An iterative evaluation of the size, power supply and cooling system requirements of these components allows the judgment and the preliminary design optimization on the considered reactor concept. The versatility of the code allows its application both to next generation tokamak devices and power reactor concepts

  19. At-reactor storage concepts criteria for preliminary assessment

    International Nuclear Information System (INIS)

    Boydston, L.A.

    1981-12-01

    The licensing, safety, and environmental considerations of four wet and four dry at-reactor storage concepts are presented. Physical criteria for each concept are examined to determine the minimum site and facility requirements which must be met by a utility which desires to expand its at-reactor spent fuel storage capability

  20. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  1. Preliminary Design Concept for a Reactor-internal CRDM

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

    2013-01-01

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

  2. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  3. In-vessel maintenance concepts for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Yount, J.A.

    1983-01-01

    Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

  4. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs

  5. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

    Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

    1989-01-01

    In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

  6. Status of fusion reactor concept development in Japan

    International Nuclear Information System (INIS)

    Tsuji-Iio, Shunji

    1996-01-01

    Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

  7. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  9. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  10. Development of a membrane-assisted fluidized bed reactor - 2 - Experimental demonstration and modeling for the partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Laverman, J.A.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    A small laboratory-scale membrane-assisted fluidized bed reactor (MAFBR) was constructed in order to experimentally demonstrate the reactor concept for the partial oxidation of methanol to formaldehyde. Methanol conversion and product selectivities were measured at various overall fluidization

  11. Alternative fusion concepts and the prospects for improved reactors

    International Nuclear Information System (INIS)

    Krakowski, R.A.

    1985-01-01

    Past trends, present status, and future directions in the search for an improved fusion reactor are reviewed, and promising options available to boh the principle tokamak and other supporting concept are summarized

  12. The Integral Fast Reactor (IFR) concept

    International Nuclear Information System (INIS)

    Till, C.E.; Chang, Y.I.

    1989-01-01

    In addition to maintaining the viability of its present commercial nuclear technology, a principal challenge in the US in the 1990s and beyond will be to regain and maintain a position among the world leadership in advanced reactor research and development. In this paper we'll discuss factors which we believe should today provide the rationale and focus for advanced reactor R and D, and we will then review the status of the major US effort, the Integral Fast Reactor (IFR) program

  13. The problems of thermohydraulics of prospective fast reactor concepts

    International Nuclear Information System (INIS)

    Sedov, A.A.

    2000-01-01

    In this report the main requirements to fast reactors in system of future multicomponent Nuclear Power with closed U-Pu fuel cycle are regarded. The peculiarities of different liquid-metal (sodium and lead-alloyed) coolants as well as the thermohydraulics problems of prospective fast reactors (FR) concepts are discussed. (author)

  14. Evaluation of Metal-Fueled Surface Reactor Concepts

    International Nuclear Information System (INIS)

    Poston, David I.; Marcille, Thomas F.; Kapernick, Richard J.; Hiatt, Matthew T.; Amiri, Benjamin W.

    2007-01-01

    Surface fission power systems for use on the Moon and Mars may provide the first use of near-term reactor technology in space. Most near-term surface reactor concepts specify reactor temperatures <1000 K to allow the use of established material and power conversion technology and minimize the impact of the in-situ environment. Metal alloy fuels (e.g. U-10Zr and U-10Mo) have not traditionally been considered for space reactors because of high-temperature requirements, but they might be an attractive option for these lower temperature surface power missions. In addition to temperature limitations, metal fuels are also known to swell significantly at rather low fuel burnups (∼1 a/o), but near-term surface missions can mitigate this concern as well, because power and lifetime requirements generally keep fuel burnups <1 a/o. If temperature and swelling issues are not a concern, then a surface reactor concept may be able to benefit from the high uranium density and relative ease of manufacture of metal fuels. This paper investigates two reactor concepts that utilize metal fuels. It is found that these concepts compare very well to concepts that utilize other fuels (UN, UO2, UZrH) on a mass basis, while also providing the potential to simplify material safeguards issues

  15. Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

    International Nuclear Information System (INIS)

    Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

    1981-09-01

    The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized

  16. Small ex-core heat pipe thermionic reactor concept (SEHPTR)

    International Nuclear Information System (INIS)

    Jacox, M.G.; Bennett, R.G.; Lundberg, L.B.; Miller, B.G.; Drexler, R.L.

    1991-01-01

    The Idaho National Engineering Laboratory (INEL) has developed an innovative space nuclear power concept with unique features and significant advantages for both Defense and Civilian space missions. The Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR) concept was developed in response to Air Force needs for space nuclear power in the range of 10 to 40 kilowatts. This paper describes the SEHPTR concept and discusses the key technical issues and advantages of such a system

  17. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  18. Advanced Nuclear Reactor Concepts for China

    International Nuclear Information System (INIS)

    Knoche, D.; Sassen, F.; Tietsch, W.; Yujie, Dong; Li, Cao

    2008-01-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  19. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  20. Safety characteristics of the integral fast reactor concept

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

    1985-01-01

    The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

  1. Preliminary design concepts for the advanced neutron source reactor systems

    International Nuclear Information System (INIS)

    Peretz, F.J.

    1988-01-01

    This paper describes the initial design work to develop the reactor systems hardware concepts for the advanced neutron source (ANS) reactor. This project has not yet entered the conceptual design phase; thus, design efforts are quite preliminary. This paper presents the collective work of members of the Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc., Engineering Division, and other participating organizations. The primary purpose of this effort is to show that the ANS reactor concept is realistic from a hardware standpoint and to show that project objectives can be met. It also serves to generate physical models for use in neutronic and thermal-hydraulic core design efforts and defines the constraints and objectives for the design. Finally, this effort will develop the criteria for use in the conceptual design of the reactor

  2. Utilization of particle fuels in different reactor concepts

    International Nuclear Information System (INIS)

    1983-04-01

    To date, particle fuel is only used in high temperature reactors (HTR). In this reactor type the particles exist of oxide fuel with a diameter of about 0.5 mm and are surrounded by various coatings in order to safely enclose fission products and decrease the radioactive release into the primary circuit. However, it is felt that fuel based upon spherical particles could have some advantages compared with pellets both on fabrication and in-core behaviour in several reactor concepts. This fuel is now of general interest and there is a high level of research and development activity in some countries. In order to collect, organize additional information and summarize experience on utilization of particle fuels in different reactor concepts, a questionnaire was prepared by IAEA in 1980 and sent to Member States, which might be involved in relevant developments. This survey has been prepared by a group of consultants and is mainly based on the responses to the IAEA questionnaire

  3. Fuel transfer cask concept design for reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Ahmad Nabil Ab Rahim; Phongsakorn Prak; Tonny Lanyau; Mohd Fazli Zakaria

    2010-01-01

    Reactor Triga PUSPATI (RTP) has been operated since 1982 till now. For such long period, the organization feels the need to upgrade the power from 1 MW to 3 MW which involved changing new fuels. Spent fuels will be stored in a Spent Fuel Pool. The process of transferring spent fuels into Spent Fuels Pool required a fuel transfer cask. This paper discussed the design concept for the fuel transfer cast which is essential equipment for reactor upgrading mission. (author)

  4. The fixed bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sahin, S.; Sefidvash, F.

    2007-01-01

    The core of a water moderated Fixed Bed Nuclear Reactor (FBNR), possessing, for instance, an electrical power of 40 MW, consists of 1.35 million fuel pellets (9.5 t) with a diameter of 1.5 cm each. The low enriched uranium fuel is made of TRISO type microspheres used in the HTGR, embedded in a graphite matrix and cladded by a shell of 1 mm SiC. Under any thinkable operational condition the fuel temperature will be below 400 C whereas its stability limit is at about 1600 C. The first characteristic of the FBNR is, therefore, its robust fuel under relatively 'cold' operating conditions and - due to the outer SiC - shell layer - the freedom from any hydrogen production. To operate the reactor the fuel pellets are pumped by a flow of water from below into the core regions where they form a stable fixed bed of about 4 cubic meter and become critical for energy production heating the outlet water to about 330 C (at 160 bar) which feeds a steam generator. The new safety feature is now the following: In case of any abnormity (e.g. external power failure, overheating etc.) the circulating pump stops and - due to gravity - the fuel pellets fall automatically out of the core region into a helical 'fuel chamber' underneath the core where their decay heat is transferred passively by natural circulation to a water tank housing the fuel chamber. The safety principle, applied here, is: The loss of an active component (circulating pump) induces a self-controlled, passively working shut-down manoeuvre accompanied by a foolproof decay heat removal without any emergency power system or any human interaction. The fuel chamber is sealed and is transported as the only reactor component to and from the reactor site. There is no possibility to irradiate fertile fuel, too. For a long-life core (larger than a 10 years cycle time) the fuel can either be poisoned by gadolinium-oxide or by a piston type core limiter adjusting the height and controlling thereby the number of the fuel pellets in

  5. Transient Analysis Needs for Generation IV Reactor Concepts

    International Nuclear Information System (INIS)

    Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.

    2002-01-01

    The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)

  6. Reactor design concepts for radiation processing

    International Nuclear Information System (INIS)

    Berejka, A.J.

    2004-01-01

    During the formative years of irradiation processing, the 1950s and 1960s, there was laboratory and academic interest in the use of this form of energy transfer to initiate polymerization for the manufacture of plastics and in other chemical processes. Studies were often based on low-dose-rate Cobalt-60 systems. The electron beam (EB) accelerator technology of the time was not as yet at the robust and industrially reliable state that it is now at the beginning of the twenty-first century. A series of reactor designs illustrate how an electron beam can be incorporated into reactor vessels for initiating gas and liquid phase polymerizations on a continuous basis. Development of such approaches, which would rely upon contemporary, high current electron beams to initiate polymerization, would help the chemical processing industry alleviate its problems of catalyst disposal and its related environmental concerns. Systems for treating materials in bulk at low doses, such as those typically used for grain disinfection, at high through-put rates, are also illustrated. Simplified shielding is envisioned in each proposed process system

  7. [Basic concept in computer assisted surgery].

    Science.gov (United States)

    Merloz, Philippe; Wu, Hao

    2006-03-01

    To investigate application of medical digital imaging systems and computer technologies in orthopedics. The main computer-assisted surgery systems comprise the four following subcategories. (1) A collection and recording process for digital data on each patient, including preoperative images (CT scans, MRI, standard X-rays), intraoperative visualization (fluoroscopy, ultrasound), and intraoperative position and orientation of surgical instruments or bone sections (using 3D localises). Data merging based on the matching of preoperative imaging (CT scans, MRI, standard X-rays) and intraoperative visualization (anatomical landmarks, or bone surfaces digitized intraoperatively via 3D localiser; intraoperative ultrasound images processed for delineation of bone contours). (2) In cases where only intraoperative images are used for computer-assisted surgical navigation, the calibration of the intraoperative imaging system replaces the merged data system, which is then no longer necessary. (3) A system that provides aid in decision-making, so that the surgical approach is planned on basis of multimodal information: the interactive positioning of surgical instruments or bone sections transmitted via pre- or intraoperative images, display of elements to guide surgical navigation (direction, axis, orientation, length and diameter of a surgical instrument, impingement, etc. ). And (4) A system that monitors the surgical procedure, thereby ensuring that the optimal strategy defined at the preoperative stage is taken into account. It is possible that computer-assisted orthopedic surgery systems will enable surgeons to better assess the accuracy and reliability of the various operative techniques, an indispensable stage in the optimization of surgery.

  8. Knowledge gaps in economic analyses of advanced reactor concepts

    International Nuclear Information System (INIS)

    Moore, M.; Pencer, J.; Leung, L.K.H.; Sadhankar, R.

    2014-01-01

    The development of next generation nuclear systems is predicated on improvement in sustainability, safety, proliferation resistance and economics. The economic assessment of the reactor concept is required as early as in the concept development stage. The Generation IV International Forum (GIF) has developed a methodology for economic assessment of the Generation IV (GEN-IV) nuclear energy systems. The GIF economics methodology was used for the assessment of one of the reactor concepts for the Super-Critical Water-cooled Reactors (SCWR), namely the European pressure-vessel type concept referred to as the High Performance Light Water Reactor (HPLWR). The economic analysis involved studying the sensitivity of two main economic indicators, namely, the Levelized Unit Electricity Cost (LUEC) and the Total Capital Investment Cost (TCIC). The knowledge gaps in estimating the capital costs and fuel costs, as well as the uncertainties in other cost parameters affecting the economic assessment of the nuclear energy system in the concept development stage are presented. (author)

  9. Advanced Burner Reactor 1000MWth Reference Concept

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kellogg, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Salev, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Seidensticker, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Tang, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Tzanos, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Chikazawa, Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2007-09-30

    The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence, to validate the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat.

  10. Aqueous self-cooled blanket concepts for fusion reactors

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1987-01-01

    A novel aqueous self-cooled blanket (ASCB) concept has been proposed. The water coolant also serves as the tritium breeding medium by dissolving small amounts of lithium compound in the water. The tritium recovery requirements of the ASCB concept may be facilitated by the novel in-situ radiolytic tritium separation technique in development at Chalk River Nuclear Laboratories. In this separation process deuterium gas is bubbled through the blanket coolant. Due to radiation induced processes, the equilibrium constant favors tritium migration to the deuterium gas stream. It is expected that the inherent simplicity of this design will result in a highly reliable, safe and economically attractive breeding blanket for fusion reactors. The available base of relevant information accumulated through water-cooled fission reactor programs should greatly facilitate the R and D effort required to validate the proposed blanket concept. Tests for tritium separation and corrosion compatibility show encouraging results for the feasibility of this concept

  11. Description of reactor fuel breeding with three integral concepts

    International Nuclear Information System (INIS)

    Ott, K.O.; Hanan, N.A.; Maudlin, P.J.; Borg, R.C.

    1979-01-01

    The time-dependent breeding of fuel in a growing system of breeder reactors can be characterized by the transitory (instantaneous) growth rate, γ(t). The three most important aspects of γ(t) can be expressed by time-independent integral concepts. Two of these concepts are in widespread use. A third integral concept that links the two earlier ones is introduced. The time-dependent growth rate has an asymptotic value, γ/sup infinity/, the equilibrium growth rate, which is the basis for the calculation of the doubling time. The equilibrium growth rate measures the breeding capability and represents a reactor property. Maximum deviation of γ(t) and γ/sup infinity/ generally appears at the initial startup of the reactor, where γ(t = 0) = γ 0 . This deviation is due to the difference between the initial and asymptotic fuel inventory composition. The initial growth rate can be considered a second integral concept; it characterizes the breeding of a particular fuel in a given reactor. Growth rates are logarithmic derivatives of the growing mass of fuel in breeder reactors, especially γ/sup infinity/, which describes the asymptotic growth by exp(γ/sup infinity/t). There is, however, a variation in the fuel-mass factor in front of this exponential function during the transition from γ 0 to γ/sup infinity/. It is shown that this variation of the fuel mass during transitioncan be described by a third integral concept, termed the breeding bonus, b. The breeding bonus measures the quality of a fuel for its use in a given reactor in terms of its impact on the magnitude of the asymptotically growing fuel mass. The calculation of γ 0 and γ/sup infinity/ is facilitated by use of the critical mass (CM) worths and the breeding worth factors, respectively

  12. Break preclusion concept and its application to the EPRTM reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chapuliot, S., E-mail: stephane.chapuliot@areva.com; Migné, C.

    2014-04-01

    This paper provides a synthesis of the technical basis supporting the break preclusion concept and its implementation on the Main Coolant Lines and Main Steam Lines of the EPR™ reactor. In a first step, it describes the background of the break preclusion concept, and then it details the requirements associated to its implementation in a Defense In Depth approach.In second steps, main benefits and few illustrative examples are given for the MCL.

  13. Nuclear-reactor remote-monitoring systems - concepts and implementations

    International Nuclear Information System (INIS)

    Rudolf, A.

    1987-01-01

    The paper presents general concepts and some examples of implemented nuclear-reactor remote-monitoring (RM) systems. Some functions and tasks of RM systems are demonstrated and three concepts are described in detail and assessed globally. Three examples of implemented RM systems are discussed using the Baden-Wurttemberg RM system for a description in greater detail. A brief prognosis of the future development of RM systems is made. (orig./DG) [de

  14. Evaluation of the breed/burn fast reactor concept

    International Nuclear Information System (INIS)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH 16 ) as the moderator

  15. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  16. Applying chemical engineering concepts to non-thermal plasma reactors

    Science.gov (United States)

    Pedro AFFONSO, NOBREGA; Alain, GAUNAND; Vandad, ROHANI; François, CAUNEAU; Laurent, FULCHERI

    2018-06-01

    Process scale-up remains a considerable challenge for environmental applications of non-thermal plasmas. Undersanding the impact of reactor hydrodynamics in the performance of the process is a key step to overcome this challenge. In this work, we apply chemical engineering concepts to analyse the impact that different non-thermal plasma reactor configurations and regimes, such as laminar or plug flow, may have on the reactor performance. We do this in the particular context of the removal of pollutants by non-thermal plasmas, for which a simplified model is available. We generalise this model to different reactor configurations and, under certain hypotheses, we show that a reactor in the laminar regime may have a behaviour significantly different from one in the plug flow regime, often assumed in the non-thermal plasma literature. On the other hand, we show that a packed-bed reactor behaves very similarly to one in the plug flow regime. Beyond those results, the reader will find in this work a quick introduction to chemical reaction engineering concepts.

  17. Reversed-Field Pinch Reactor (RFPR) concept

    International Nuclear Information System (INIS)

    Hagenson, R.L.; Krakowski, R.A.; Cort, G.E.

    1979-08-01

    A conceptual engineering design of a fusion reactor based on plasma confinement in a Reversed-Field Pinch (FRP) configuration is presented. A 50% atomic mixture of deuterium and tritium (DT) is ohmically heated to ignition by currents flowing in the toroidal plasma; this plasma current also inherently produces the confining magnetic fields in a toroidal chamber having a major and minor radii of 12.7 and 1.5 m, respectively. The DT plasma ignites in 2 to 3 s and burns at 10 to 20 keV for approx. 20 s to give a fuel burnup of approx. 50%. Tritium breeding occurs in a granular Li 2 O blanket which is packed around an array of radially oriented coolant tubes carrying a mixture of high-pressure steam and water. The slightly superheated steam emerging from this blanket would be used to drive a turbine directly. Low-pressure helium containing trace amounts of oxygen is circulated through the packed Li 2 O bed to extract the tritium. A 20-mm-thick copper first wall serves as a neutron multiplier, acts as a tritium barrier, and supports image currents to provide plasma stabilization on a 0.1-s timescale; external windings provide stability for longer times

  18. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  19. Axial heterogeneous core concept applied for super phoenix reactor

    International Nuclear Information System (INIS)

    Batista, J.L.; Renke, C.A.C.; Waintraub, M.; Santos Bastos, W. dos; Brito Aghina, L.O. de.

    1991-11-01

    Always maintaining the current design rules, this paper presents a parametric study on the type of axial heterogeneous core concept (CHA), utilizing a core of fast reactor Super Phenix type, reaching a maximum thermal burnup rate of 150000 M W d/t and being managed in single batch. (author)

  20. Concept of innovative water reactor for flexible fuel cycle (FLWR)

    International Nuclear Information System (INIS)

    Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

    2005-01-01

    In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

  1. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Park, J.Y.; Ruther, W.E.; Kassner, T.F.; Shack, W.J.

    1990-12-01

    Topics that have been investigated during this year include (1) SCC of A533-Gr B steel used in steam generator and reactor pressure vessels, (2) fatigue of Type 316NG SS, and (3) SCC of Type 347 and CF-3 cast duplex stainless steels in simulated BWR water. Crack-growth-rate (CGR) tests were performed on a composite A533-Gr B/Inconel-182 specimen in which the stress corrosion crack in the Inconel-182 weld metal penetrated and grew into the A533-Gr B steel. CGR tests were also conducted on conventional (unplated) and nickel- or gold-plated A533-Gr B specimens to provide insight into whether the nature of the surface layer on the low-alloy steel, either oxide corrosion products or a noble metal, influences the overall SCC process. CGR data on the A533-Gr B specimens were compared with the fatigue crack reference curves in the ASME Boiler and Pressure Vessel Code, Section XI, Appendix A. Fatigue tests were conducted on Type 316NG SS in air and simulated BWR water at low strain ranges and frequencies to better establish margins in the ASME Code Section III Fatigue Design Curves. CGR tests were also conducted on specimens of Type 347 SS with different heat-treatment conditions, and a specimen of CF-3 cast stainless steel with a ferrite content of 15.6%. The results were compared with previous data on another heat of Type 347 SS, which was very resistant to SCC, and a CF-3M steel with a ferrite content of 5%. 37 refs., 15 figs., 8 tabs

  2. A novel concept for CRIEC-driven subcritical research reactors

    International Nuclear Information System (INIS)

    Nieto, M.; Miley, G.H.

    2001-01-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  3. Comparative analysis of sub-critical transmutation reactor concepts

    International Nuclear Information System (INIS)

    Chang, S. H.

    1997-01-01

    The long-lived nuclear wastes have been substantially generated from the light water reactor for a few decades. The toxicity of these spent fuels will be higher than that of the uranium ore, even if those will be stored in the repository more than ten thousands. Hence the means of transmuting the key long-lived nuclear wastes, primarily the minor actinides, using a hybrid proton accelerator and subcritical transmutation reactor, are proposed. Until now, the representative concepts for a subcritical transmutation reactor are the Energy Amplifier, the OMEGA project, the ATW and the MSBR. The detailed concepts and the specifications are illustrated in Table 1. The design requirements for the subcritical transmutation reactor are the high transmutation rate of long-lived nuclear wastes, safety and economics. And to propose the subcritical transmutation reactor concepts, the coolant, the target material and fuel type are carefully considered. In these aspects, the representative concepts for a subcritical transmutation reactor in Table 1 have been surveyed. The requirements for a target and a coolant are the reliable, low maintenance operation and safe operation to minimize the wastes. The reliable, low maintenance operation and safe operation to minimize the wastes. The reliable coolant must have the low melting point, high heat capacity and excellent physical properties. And the target material must have high neutron yield for a given proton condition and easy heat removal capability. Therefore in respect with the above requirements, Pb-Bi is proposed as the coolant and the target material for the subcritical reactor. Because the neutron yield for a given proton energy increases linearly with mass number up to bismuth but in heavier elements spallation events sharply increase both the neutron and heat outputs, Pb-Bi meets not only such the requirements as the above for the coolant but also those for the coolant and target, the simplification of system can be achieved

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

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

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

  5. Decommissioning of the AVR reactor, concept for the total dismantling

    International Nuclear Information System (INIS)

    Marnet, C.; Wimmers, M.; Birkhold, U.

    1998-01-01

    After more than 21 years of operation, the 15 MWe AVR experimental nuclear power plant with pebble bed high temperature gas-cooled reactor was shout down in 1988. Safestore decommissioning began in 1994. In order to completely dismantle the plant, a concept for Continued dismantling was developed according to which the plant could be dismantled in a step-wise procedure. After each step, there is the possibility to transform the plant into a new state of safe enclosure. The continued dismantling comprises three further steps following Safestore decommissioning: 1. Dismantling the reactor vessels with internals; 2. Dismantling the containment and the auxiliary units; 3. Gauging the buildings to radiation limit, release from the validity range of the AtG (Nuclear Act), and demolition of the buildings. For these steps, various technical procedures and concepts were developed, resulting in a reference concept in which the containment will essentially remain intact (in-situ concept). Over the top of the outer reactor vessel a disassembling area for remotely controlled tools will be erected that tightens on that vessel and can move down on the vessel according to the dismantling progress. (author)

  6. State of the art of the fluidized bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sefidvash, F.; Vilhena, M.T.M.B. de; Streck, E.; Borges, V.; Johansson, M.

    1987-01-01

    A small and simple nuclear reactors with inherent safety using the fluidized bed concept is under research and study. In this paper a brief study neutronics and thermal hydraulics of this reactor concept is presented. (Author) [pt

  7. The multi region molten-salt reactor concept

    International Nuclear Information System (INIS)

    Gyula, Csom; Sandor, Feher; Szieberth, M.; Szabolcs, Czifrus

    2003-01-01

    The molten-salt reactor (MSR) concept is one of the most promising systems for the realisation of transmutation. The objective is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures. The procedure is the multi-step transmutation, in which the transformation is carried out in several consecutive steps of different neutron flux and spectrum. In order to implement this, a multi-region transmutation device, i.e. nuclear reactor or sub-critical system is proposed, in which several separate flow-through irradiation rooms are formed with various neutron spectra and fluxes. The paper presents calculations that were performed for a special 5-region version of the multi-region molten-salt reactor. (author)

  8. Analysis of a sustainable gas cooled fast breeder reactor concept

    International Nuclear Information System (INIS)

    Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

    2014-01-01

    Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

  9. Identification of improvements of advanced light water reactor concepts

    International Nuclear Information System (INIS)

    Frisch, W.; Liesch, K.; Riegel, B.

    1993-01-01

    The scope of this report is to identify the improvement of reactor developments with respect to reactor safety. This includes the collection of non-proprietary information on the description of the advanced design characteristics, especially summary design descriptions and general publications. This documentation is not intended to include a safety evaluation of the advanced concepts; however, it is structured in such a way that it can serve as a basis for a future safety evaluation. This is taken into account in the structure of the information regarding the distinction of the various concepts with respect to their 'advancement' and the classification of design characteristics according to some basic safety aspects. The overall description concentrates on those features which are relevant to safety. Other aspects, such as economy, operational features, maintenance, the construction period, etc...are not considered explicitly in this report

  10. Concept design on RH maintenance of CFETR Tokamak reactor

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Songtao; Wan, Yuanxi; Li, Jiangang; Ye, Minyou; Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua

    2014-01-01

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed

  11. Innovative Control concepts for German pressurized water reactors

    International Nuclear Information System (INIS)

    Brzozowski, Raphael; Kuhn, Andreas

    2010-01-01

    Controlling reactor power without any manual support is becoming more and more important. The READIG project (READIG = Reactor Instrumentation and Digital Control) power control system installed in unit 2 of the Philippsburg nuclear power station (KKP 2) requires no manual intervention except for specific strategy criteria settings. It was even possible to eliminate the power distribution set points. With minor adaptations, this concept can be applied in other PWR plants as well. KKP 2 is a PWR plant with particularly sophisticated core charges; as a consequence, the I and C systems were adapted accordingly. The increase in integral reactor power and the low-leakage core charges are the main reasons for lower limiting margins, especially in peak limiting. The standard control concept was supplemented in such a way that a more precise fine control concept for power distribution in the full-load regime is achieved. The READIG project fully utilizes the possibilities offered by digital TXS Technology, which is why use is also made of physical parameterization. The new power distribution control concept has these advantages: - Operation at small peak-/DNB-reactor output limitation margins. - Stable control without manual intervention also in load cycles and in the frequency control mode. - Simplified operation due to omission of the power distribution set point. - Reduction to zero of the frequency of L-bank steps at constant power with superimposed frequency control mode. - Reduction to zero of the frequency of D-bank steps at constant power with superimposed frequency control mode. - Lower quantities of demineralized water to be fed at constant power with superimposed frequency control mode (±1%). (orig.)

  12. Steady-state operation requirements of tokamak fusion reactor concepts

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

    In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

  13. Environmentally assisted cracking of light-water reactor materials

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

    1996-02-01

    Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used

  14. New concepts for controlled fusion reactor blanket design

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Avci, H.; El-Maghrabi, M.

    1975-01-01

    Several new concepts for fusion reactor blanket design based on the idea of shifting, or tailoring, the neutron spectrum incident on the first structural wall are presented. The spectral shifter is a nonstructural element which can be made of graphite, silicon carbide, or three dimensionally woven carbon fibers (and containing other materials as appropriate) placed between the neutron source and the first structural wall. The softened neutron spectrum incident on the structural components leads to lower gas production and atom displacement rates than in more standard fusion blanket designs. In turn, this results in longer anticipated lifetimes for the structural materials and can significantly reduce radioactivity and afterheat levels. In addition, the neutron spectrum in the first structural wall can be made to approach the flux shape in fast breeder reactors. Such spectral softening means that existing radiation facilities may be more profitably used to provide relevant materials radiation damage data for the structural materials in these fusion blanket designs. This general class of blanket concepts are referred to as internal spectral shifter and energy converter, or ISSEC concepts. These specific design concepts fall into three main categories: ISSEC/EB concepts based on utilizing existing designs which breed tritium behind the first structural wall; ISSEC/IB concepts based on breeding tritium inside the first vacuum wall; and ISSEC/Bu concepts based on using boron, carbon, and perhaps, beryllium to obtain an energy multiplier and converter design that does not attempt to breed tritium or utilize lithium. The detailed analyses relate specifically to the nuclear performance of ISSEC systems and to a discussion of materials radiation damage problems in the structural material.(U.S.)

  15. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  16. Concept study of the Steady State Tokamak Reactor (SSTR)

    International Nuclear Information System (INIS)

    1991-06-01

    The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  18. 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

  19. Study on plant concept for gas cooled fast reactor

    International Nuclear Information System (INIS)

    Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

    2003-05-01

    In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

  20. Concept of spatial channel theory applied to reactor shielding analysis

    International Nuclear Information System (INIS)

    Williams, M.L.; Engle, W.W. Jr.

    1977-01-01

    The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield

  1. Conception of divertorless tokamak reactor with turbulent plasma blanket

    International Nuclear Information System (INIS)

    Nedospasov, A.V.; Tokar, M.Z.

    1980-01-01

    The results of the calculations presented here demonstrate that, with technically reasonable degree of the magnetic field stochastisation, the turbulent plasma blanket can take the place of a divertor. It performs the three main functions of the divertor: (a) the exhaust of the helium and unburned fuel; (b) weakening of the fast particle flux to the wall surface; and (c) essential reduction of the impurity content in the active zone of the reactor. Taking into account that plasma flows to the first wall along field lines, we may figuratively say that the first wall plays the role of a divertor in our conception. (orig.)

  2. Technical feasibility study of 60 MWe fast reactor concept: RAPID

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

    1993-01-01

    A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

  3. Fast reactor core concepts to improve transmutation efficiency

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi

    2015-01-01

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate

  4. Preliminary concepts: safeguards for spent light-water reactor fuels

    International Nuclear Information System (INIS)

    Cobb, D.D.; Dayem, H.A.; Dietz, R.J.

    1979-06-01

    The technology available for safeguarding spent nuclear fuels from light-water power reactors is reviewed, and preliminary concepts for a spent-fuel safeguards system are presented. Essential elements of a spent-fuel safeguards system are infrequent on-site inspections, containment and surveillance systems to assure the integrity of stored fuel between inspections, and nondestructive measurements of the fuel assemblies. Key safeguards research and development activities necessary to implement such a system are identified. These activities include the development of tamper-indicating fuel-assembly identification systems and the design and development of nondestructive spent-fuel measurement systems

  5. MAPLE: a Canadian multipurpose reactor concept for national nuclear development

    International Nuclear Information System (INIS)

    Lidstone, R.F.

    1984-06-01

    Atomic Energy of Canada Limited, following an investigation of Canadian and international needs and world-market prospects for research reactors, has developed a new multipurpose concept, called MAPLE (Multipurpose Applied Physics Lattice Experimental). The MAPLE concept combines H 2 O- and D 2 O-moderated lattices within a D 2 O calandria tank in order to achieve the flux advantages of a basic H 2 O-cooled and moderated core along with the flexibility and space of a D 2 O-moderated core. The SUGAR (Slowpoke Uprated for General Applied Research) MAPLE version of the conept provides a range of utilization that is well suited to the needs of countries with nuclear programs at an early stage. The higher power MAPLE version furnishes high neutron flux levels and the variety of irradiation facilities that are appropriate for more advanced nuclear programs

  6. Plant Control Concept for the Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, S. O.

    2010-12-01

    A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

  7. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  8. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  9. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  10. Strategy-oriented display concept to assist severe accident management

    International Nuclear Information System (INIS)

    Jeong, Kwangsub; Ha, Jaejoo

    2000-01-01

    The Critical Function Monitoring System (CFMS) is a typical Safety Parameter Display System (SPDS) to assist the operation of Korean Standard Nuclear Power Plants during normal and emergency operation, and SPDS for severe accident is being developed in Korea. When the existing CFMS is used under a severe accident situation, some problems are expected from: (1) different design basis, i.e. prevention of core melt vs. protection of radiation release to environment, (2) different parameters for decision-making, and (3) different domain and depth of information to restore the plant. To resolve the above problems, a concept, 'Strategy-Oriented Information Display' concept, for displaying information for severe accident management is developed in this paper. Whereas the existing SPDS structure is based on the critical safety function, the developed concept is based on the severe accident management strategy. The display for each strategy includes the plant parameters to check the status of plant and component with the logical or graphical views necessary for executing the strategy. As the application of the proposed concept, KAERI is developing a display system, the prototype severe accident SPDS, Severe Accident Management Display System (SAMDIS), to assist plant personnel for executing Korean Severe Accident Management Guidelines. CFMS is developed for a general display suitable to all situations with various displays. On the contrary, SAMDIS provides all the relevant information on one screen based on the proposed concept. The SAMDIS screen shows more extensive area than CFMS and thus plant personnel can recognize the overall plant status at a glance. This concept is quite effective when used with severe accident management guidelines because of the relatively macroscopic characteristics of a severe accident management strategy. (author)

  11. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  12. Evolution of the technical concept of fast reactors. The concept of BREST

    International Nuclear Information System (INIS)

    Orlov, V.V.

    2001-01-01

    Having understood that conventional power was limited by available fuel resources, as well as the environmental concern, and willing to use the advantages of defense nuclear power achievements, the development of civil nuclear power was initiated. Scarce supply of uranium has been a matter of concern from the very beginning of nuclear power development, but plutonium produced in the thermal reactors was supposed to be used as fuel for the fast reactors which would not be limited by fuel resources. In order to attain high breeding ratio and high power density, the first generation of fast reactors were designed with sodium coolant, uranium blanket to make up for a decrease in breeding ratio if uranium oxides were used as fuel. Development of nuclear power in the sixties and seventies was followed by stagnation. Lessons learned from a 50-year experience and new conditions set for power industry demand a new concept of fast reactor which would meet a variety of cost-efficiency and safety requirements in their present understanding. Development of fast breeders in Russia began after commissioning of BN-350 and completion of BN-600 design. According to present demands BREST reactors should be designed so as to implement consistently the principles of natural safety without deviation from materials and technology which was proven in defense and civil nuclear power facilities

  13. 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)

  14. Old and new ways in reactor technology. Reactor concepts and reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Schulten, R

    1989-01-01

    Compared to developments of other technical-scale systems, the period between the recognition of the underlying physics of nuclear fission and the development of a functioning nuclear reactor and its further development to the present level of maturity has been relatively short. The whole development is based on the chain reaction and is rendered safe by the possible auto-stabilization of this reaction. Consequently, the safety of nuclear reactors properly designed is based on automatic mechanisms, which prevent spreads of radioactivity even in major accidents. Controversial opinions about nuclear power uses are mostly based on wrong perceptions both of reactor safety and of radioactive waste, unless they are characterized by sheer ideology. The use of nuclear power worldwide has assumed an important, growing role in the combined uses of a variety energy sources in a surprisingly short period of time and will continue to make a safe, economic, and thus responsible contribution in the long run.

  15. PX–An Innovative Safety Concept for an Unmanned Reactor

    Directory of Open Access Journals (Sweden)

    Sung-Jae Yi

    2016-02-01

    Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

  16. Experimental investigation of the MSFR molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-15

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

  17. High density, high magnetic field concepts for compact fusion reactors

    International Nuclear Information System (INIS)

    Perkins, L.J.

    1996-01-01

    One rather discouraging feature of our conventional approaches to fusion energy is that they do not appear to lend themselves to a small reactor for developmental purposes. This is in contrast with the normal evolution of a new technology which typically proceeds to a full scale commercial plant via a set of graduated steps. Accordingly' several concepts concerned with dense plasma fusion systems are being studied theoretically and experimentally. A common aspect is that they employ: (a) high to very high plasma densities (∼10 16 cm -3 to ∼10 26 cm -3 ) and (b) magnetic fields. If they could be shown to be viable at high fusion Q, they could conceivably lead to compact and inexpensive commercial reactors. At least, their compactness suggests that both proof of principle experiments and development costs will be relatively inexpensive compared with the present conventional approaches. In this paper, the following concepts are considered: (1) The staged Z-pinch, (2) Liner implosion of closed-field-line configurations, (3) Magnetic ''fast'' ignition of inertial fusion targets, (4) The continuous flow Z-pinch

  18. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  19. Lawson concepts and criticality in DT fusion reactors

    International Nuclear Information System (INIS)

    Lartigue, J.G.

    1987-01-01

    The original Lawson concepts (amplification factor R and parameter nτ) as well as their applications in DT reactors are discussed in two cases: the ignition regime and the subignition regime in a self-sufficient plant. The modified Lawson factor or internal amplification factor R a (a function of alpha power) is proposed as a means to measure the ignition level reached by the plasma, in a more precise way than that given by the collective parameter (nτkT). The self-sufficiency factor (δ) is proposed as a means to measure the plant self-sufficiency, δ being more significant than the traditional Q factor. It is stated that the ignition regime (R a = 1) is equivalent to a critical state (energy equilibrium); then, the corresponding critical mass concept is proposed. The analysis of the R a relationship with temperature (kT), (nτ), and recirculating factor (var-epsilon) gives the conditions for the reactor to reach ignition or for the plant to reach self-sufficiency; it also shows that an approach to ignition is not improved by heating from 50 to 100 KeV

  20. The dual fluid reactor - a new concept for a highly effective fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huke, A.; Ruprecht, G. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); WeiBbach, D. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. Szczecin, ul. Wielkopolska, Inst. Fizyki, Wydzial Matematyczno-Fizyczny, Szczecin, (Poland); Gottlieb, S. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Hussein, A. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. of Northern British Columbia, Dept. of Physics, Prince George, BC (Canada); Czerski, K. [Institut fur Festkorper-Kernphysik gGmbH, Leistikowstr, Berlin (Germany); Univ. Szczecin, ul. Wielkopolska, Inst. Fizyki, Wydzial Matematyczno-Fizyczny, Szczecin, (Poland)

    2014-07-01

    The Dual Fluid Reactor, DFR, is a novel concept of a fast heterogeneous nuclear reactor. Its key feature is the employment of two separate liquid cycles, one for fuel and one for the coolant. As opposed to other liquid-fuel concepts like the molten-salt fast reactor (MSFR), in the DFR both cycles can be separately optimized for their respective purpose, leading to advantageous consequences: A very high power density resulting in enormous cost savings, and a highly negative temperature feedback coefficient, enabling a self-regulation without any control rods or mechanical parts in the core. The fuel liquid, an undiluted actinide trichloride (consisting of isotope-purified {sup 37}Cl) in the reference design, circulates at an operating temperature of 1300 K and can be processed on-line in a small internal processing unit utilizing fractionated distillation or electro refining. Medical radioisotopes like Mo-99/Tc-99m are by-products and can be provided right away. In a more advanced design, an actinide metal alloy melt with an appropriately low solidus temperature is well possible which further compactifies the core and allows to further increase the operating temperature due to its high heat conductivity. The best choice for the coolant is pure lead which yields a very hard neutron spectrum. (author)

  1. REACTOR - a Concept for establishing a System-of-Systems

    Science.gov (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    well suited to establish brokers, which mediate metadata and semantic information about the resources of all involved systems. This concept has been developed within the project Collaborative, Complex, and Critical Decision-Support in Evolving Crises (TRIDEC) on the basis of semantic registries describing all facets of events and services utilisable for crisis management systems. The implementation utilises an operative infrastructure including an Enterprise Service Bus (ESB), adapters to proprietary sensor systems, a workflow engine, and a broker-based MOM. It also applies current technologies like actor-based frameworks for highly concurrent, distributed, and fault tolerant event-driven applications. Therefore REACTOR implementations are well suited to be hosted in a cloud that provides Infrastructure as a Service (IaaS). To provide low entry barriers for legacy and future systems, REACTOR adapts the principles of Design by Contract (DbC) as well as standardised and common information models like the Sensor Web Enablement (SWE) or the JavaScript Object Notation for geographic features (GeoJSON). REACTOR has been applied exemplarily within two different scenarios, Natural Crisis Management and Industrial Subsurface Development.

  2. Summary of Research on Light Water Reactor Improvement Concepts

    International Nuclear Information System (INIS)

    Mowery, Alfred L.

    2002-01-01

    The Arms Control and Disarmament Agency of the U.S. Department of State instituted a study aimed at improving the light water reactor (LWR) fuel consumption efficiency as an alternative to fuel recycle in the late 1970s. Comparison of the neutron balance tables of an LWR (1982 design) and an 'advanced' Canada deuterium uranium (CANDU) reactor explained that the relatively low fuel efficiency of the LWR was not primarily a consequence of water moderator absorptions. Rather, the comparatively low LWR fuel efficiency resulted from its use of poison to hold down startup reactivity together with other neutron losses. The research showed that each neutron saved could reduce fuel consumption by about 5%. In a typical LWR some 5 neutrons (out of 100) were absorbed in control poisons over a cycle. There are even more parasitic and leakage neutron absorptions. The objective of the research was to find ways to minimize control, parasitic, and other neutron losses aimed at improved LWR fuel consumption. Further research developed the concept of 'putting neutrons in the bank' in 238 U early in life and 'drawing them out of the bank' late in life by burning the 239 Pu produced. Conceptual designs were explored that could both control the reactor and substantially improve fuel efficiency and minimize separative work requirements.The U.S. Department of Energy augmented its high burnup fuel program based on the research in the late 1970s. As a result of the success of this program, fuel burnup in U.S. LWRs has almost doubled in the intervening two decades

  3. Blanket concepts for the ARIES commercial tokamak reactor study

    International Nuclear Information System (INIS)

    Grotz, S.P.; Ghoniem, N.M.; Hasan, M.Z.; Martin, R.C.; Najmabadi, F.; Sharafat, S.; Hua, T.; Sze, D.K.; Cheng, E.T.; Creedon, R.L.; Wong, C.P.C.; Herring, J.S.; Klein, A.; Snead, L.; Steiner, D.

    1989-01-01

    The ARIES study is a 3-year effort, started in 1988, exploring the potential of the tokamak to be an attractive and competitive commercial power reactor. Several different versions of the tokamak are being considered, combining different levels of extrapolations in physics and engineering databases. The first version studied in detail, ARIES-I, combines present-day physics (with minimal extrapolation) with aggressive engineering technology such as very high-field, superconducting magnets and low-activation silicon carbide composite materials. The ARIES-I version is designed to meet acceptable safety and environmental criteria. In particular, achieving a passively safe concept that meets Class-C waste disposal is one of the high leverage items in the design. This paper summarizes the scoping analysis and engineering design of the ARIES-I fusion-power-core subsystems. The ARIES-I design is a 1000 MW e power reactor, operating at steady state in the 1 st stability regime and uses a high magnetic field. Typical operating parameters of the ARIES-I strawman design are listed

  4. Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2003-01-01

    The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

  5. Ethics and the law: the law and assisted human conception.

    Science.gov (United States)

    Brahams, D

    1990-07-01

    This article aims to provide no more than a brief summary and overview of some of the principal legal questions which arise in connection with assisted human conception. There is no requirement of legal suitability for natural parenthood, though a child may be removed from parental care at birth if its welfare is considered to be at risk. Where medical or other assistance is required, however, the law and social judgments may impinge on the freedom of individuals to procreate. Commercial surrogacy has recently been criminalized, but private surrogacy arrangements without reward are not illegal--although any contract would probably be unenforceable through the courts. If medical intervention is required to achieve assisted conception, the availability of resources for NHS treatment, the physical and mental health of the prospective mother and father, and the welfare (or lack of it) of any prospective child, may be factors in deciding whether an infertility unit will offer treatment. Such practices must not operate unfairly and must not discriminate on racial grounds. If treatment is provided, and a woman becomes pregnant, the ordinary abortion laws will apply and, it is thought, will extend to the selective reduction of a multiple pregnancy--there is no claim in English law for 'wrongful birth'. AID does not constitute adultery, and the law has recently been reformed to recognize children born following AID as legitimate to their social parents. A child may be regarded as the legitimate child of a surrogate mother's marriage, but where the baby is genetically distinct from the surrogate mother, the law, and is uncertain and as yet could be conflicting claims of parenthood without legislation. The storage and disposal of human gametes and embryos may raise problems of 'ownership'.

  6. Design features of BREST reactors. Experimental work to advance the concept of BREST reactors. Results and plans

    International Nuclear Information System (INIS)

    Filin, A.I.; Orlov, V.V.; Leonov, V.N.; Sila-Novitskij, A.G.; Smirnov, V.S.; Tsikunov, V.S.

    2001-01-01

    Principle designs of 300 MW(th) and 1200 MW(th) lead-cooled fast reactors are presented. Reactors of various output are shown to be built using the same principles. In conjunction with increased output and to implement inherent safety concept in BREST-1200 reactor design a number of new solutions, which may be used in BREST-300 concept too, has been taken including: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using Field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by-pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

  7. Antenatal mood and fetal attachment after assisted conception.

    Science.gov (United States)

    Fisher, Jane R W; Hammarberg, Karin; Baker, Gordon H W

    2008-05-01

    Australian women conceiving with ART are at fourfold risk of admission to early parenting treatment programs compared with those conceiving spontaneously. This study aimed to identify prevalence and determinants of antenatal mood disturbance and other risks for early parenting difficulties after assisted conception. A prospective longitudinal investigation from conception to 18 months postpartum using telephone interviews and self-report questionnaires. Melbourne IVF and Royal Women's Hospital Reproductive Services, Victoria, Australia. A consecutive cohort of English-speaking women with ultrasound-confirmed ART-conceived pregnancies. Standardized psychometric measures of mood, quality of marital relationship, mother-to-fetus emotional attachment, and personality. None. Of the 288 women with confirmed pregnancies, 239 were contactable, and 183 (77%) were recruited, 95% of whom completed both early and late pregnancy assessments. Participants were socioeconomically advantaged, had very good pregnancy health, exceptional marital relationships, normal personality styles, and intense affectionate attachment to the fetus. Very few (<5%) had clinically significant mood disturbance in late pregnancy. There were low rates of antenatal mood disturbance and other risk factors for postpartum depression. Pregnancy and motherhood might be idealized after ART conception, and preparation for the realities of infant care might then be insufficient.

  8. Assistant Personal Robot (APR: Conception and Application of a Tele-Operated Assisted Living Robot

    Directory of Open Access Journals (Sweden)

    Eduard Clotet

    2016-04-01

    Full Text Available This paper presents the technical description, mechanical design, electronic components, software implementation and possible applications of a tele-operated mobile robot designed as an assisted living tool. This robotic concept has been named Assistant Personal Robot (or APR for short and has been designed as a remotely telecontrolled robotic platform built to provide social and assistive services to elderly people and those with impaired mobility. The APR features a fast high-mobility motion system adapted for tele-operation in plain indoor areas, which incorporates a high-priority collision avoidance procedure. This paper presents the mechanical architecture, electrical fundaments and software implementation required in order to develop the main functionalities of an assistive robot. The APR uses a tablet in order to implement the basic peer-to-peer videoconference and tele-operation control combined with a tactile graphic user interface. The paper also presents the development of some applications proposed in the framework of an assisted living robot.

  9. Concept of Educational Assistance to Health Protection of the Individual

    Science.gov (United States)

    Levanova, Elena Aleksandrovna; Kokorina, Olga Rafailovna; Nikitin, Yuriy Vladimirovich; Perepelkina, Tatiyna Vladislavovna; Segodina, Polina Anatolievna

    2016-01-01

    The article describes the theoretical and practical need for the development of the concept of assistance to health protection of the individual in order to address the problem of health protection of students and teachers in the conditions of a higher pedagogical education. The problem of studying human health, its entirety, systemacity and connection with the environment attracts particular attention in recent years. This was one of the reasons to study the problem of “healthy lifestyle” as the qualitative characteristic of a human life aimed at health, due to the fact that a healthy lifestyle is one of the determinants of health. This is made possible with the use of specific health-protecting technologies aimed at searching for ways and means of protection and conservation of health of students and teachers in the conditions of the educational process and using educational tools, which is currently included into the priorities of education. PMID:26493439

  10. Method to assist conception of collimators in nuclear medicine

    International Nuclear Information System (INIS)

    Gantet, P.; Esquerre, J.P.; Danet, B.; Roux, G.; Guiraud, R.

    1990-01-01

    Designing a collimator should begin by an accurate computation of its performances in order to minimize the usual expensive and time consuming phase of trial and error. The authors briefly describe several methods currently used, and present a method of simulation of the percussional response of collimators. The computation takes into account the attenuation of photons by the collimator septas. An other benefit is its ability to be used whatever the geometric specifications of the collimator. The program computes spatial resolution, geometric efficiency, septal penetration, as well as slice thickness when the collimator is used with a SPECT device. The study presents the results concerning two collimators: a general purpose parallel hole commercially available and a focused parallel one dedicated to a single slice SPECT system. In conclusion, this deterministic method which takes attenuation into account for collimators performances computation should be useful tool to assist conception of new collimators in nuclear medicine [fr

  11. The hybrid reactor project based on the straight field line mirror concept

    International Nuclear Information System (INIS)

    Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestål, A.; Källne, J.; Anglart, H.

    2012-01-01

    The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with “semi-poor” plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Q r = P fis /P fus >>1. The upper bound on Q r is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Q r ≈150, corresponding to a neutron multiplicity of k eff =0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement T e ≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as

  12. The hybrid reactor project based on the straight field line mirror concept

    Science.gov (United States)

    Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestâl, A.; Källne, J.; Anglart, H.

    2012-06-01

    The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with "semi-poor" plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Qr = Pfis/Pfus>>1. The upper bound on Qr is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Qr≈150, corresponding to a neutron multiplicity of keff=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement Te≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on

  13. Calculation methods for advanced concept light water reactor lattices

    International Nuclear Information System (INIS)

    Carmona, S.

    1986-01-01

    In the last few years s several advanced concepts for fuel rod lattices have been studied. Improved fuel utilization is one of the major aims in the development of new fuel rod designs and lattice modifications. By these changes s better performance in fuel economics s fuel burnup and material endurance can be achieved in the frame of the well-known basic Light Water Reactor technology. Among the new concepts involved in these studies that have attracted serious attention are lattices consisting of arrays of annular rods duplex pellet rods or tight multicells. These new designs of fuel rods and lattices present several computational problems. The treatment of resonance shielded cross sections is a crucial point in the analyses of these advanced concepts . The purpose of this study was to assess adequate approximation methods for calculating as accurately as possible, resonance shielding for these new lattices. Although detailed and exact computational methods for the evaluation of the resonance shielding in these lattices are possible, they are quite inefficient when used in lattice codes. The computer time and memory required for this kind of computations are too large to be used in an acceptable routine manner. In order to over- come these limitations and to make the analyses possible with reasonable use of computer resources s approximation methods are necessary. Usual approximation methods, for the resonance energy regions used in routine lattice computer codes, can not adequately handle the evaluation of these new fuel rod lattices. The main contribution of the present work to advanced lattice concepts is the development of an equivalence principle for the calculation of resonance shielding in the annular fuel pellet zone of duplex pellets; the duplex pellet in this treatment consists of two fuel zones with the same absorber isotope in both regions. In the transition from a single duplex rod to an infinite array of this kind of fuel rods, the similarity of the

  14. Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

    International Nuclear Information System (INIS)

    Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

    2011-01-01

    Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

  15. Multiplicity and early gestational age contribute to an increased risk of cerebral palsy from assisted conception

    DEFF Research Database (Denmark)

    Hvidtjørn, Dorte; Grove, Jakob; Schendel, Diana

    2010-01-01

    BACKGROUND: This paper assesses the risk of cerebral palsy (CP) in children born after assisted conception compared with children born after natural conception (NC). METHODS: This population based follow-up study included all 588,967 children born in Denmark from 1995 to 2003. Assisted conception...

  16. High Flux Isotope Reactor cold neutron source reference design concept

    International Nuclear Information System (INIS)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R.

    1998-05-01

    In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

  17. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  18. The risks of nuclear energy technology. Safety concepts of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Inst. fuer Kern- und Energietechnk (IKET); Kessler, Guenter; Veser, Anke; Schlueter, Franz-Hermann

    2014-11-01

    Analyses the risks of nuclear power stations. Discusses the security concept of reactors. Analyzes possible crash of air planes on a reactor containment. Presents measures against the spread of radioactivity after a severe accident. Written in engaging style for professionals and policy makers. The book analyses the risks of nuclear power stations. The security concept of reactors is explained. Measures against the spread of radioactivity after a severe accident, accidents of core melting and a possible crash of an air plane on a reactor containment are discussed. The book covers three scientific subjects of the safety concepts of Light Water Reactors: - A first part describes the basic safety design concepts of operating German Pressurized Water Reactors and Boiling Water Reactors including accident management measures introduced after the reactor accidents of Three Mile Island and Chernobyl. These safety concepts are also compared with the experiences of the Fukushima accidents. In addition, the safety design concepts of the future modern European Pressurized Water Reactor (EPR) and of the future modern Boiling Water Reactor SWR-1000 (KERENA) are presented. These are based on new safety research results of the past decades. - In a second, part the possible crash of military or heavy commercial air planes on a reactor containment is analyzed. It is shown that reactor containments can be designed to resist to such an airplane crash. - In a third part, an online decision system is presented. It allows to analyze the distribution of radioactivity in the atmosphere and to the environment after a severe reactor accident. It provides data for decisions to be taken by authorities for the minimization of radiobiological effects to the population. This book appeals to readers who have an interest in save living conditions and some understanding for physics or engineering.

  19. Potential mirror concepts for radiation testing of fusion reactor materials

    International Nuclear Information System (INIS)

    Miley, G.H.

    1977-01-01

    Studies under the University of Illinois PROMETHEUS (Plasma Reactor Optimized for Materials Experimentation for Thermonuclear Energy Usage) project are described that started in 1971 with the realization that a practical fusion-plasma neutron source was feasible with a net-power input (rather than production). The basic objectives were similar to those in later FERF (Fusion Engineering Research Facility) studies: namely, to maximize the neutron flux and usable experimental volume; to include the flexibility to handle a variety of both materials and engineering experiments; to minimize capital and operating costs; and to utilize near- term technology. The PROMETHEUS design provides a neutron flux of approximately 5x10 14 n/cm 2 s by injection of approximately 30 MW of neutral-beams into a 20 cm radius mirror-confined plasma. Charge-exchange bombardment of the first wall is viewed as a key problem in the design and is discussed in some detail. To gain yet higher neutron fluxes for accelerated testing, two alternate designs have been studied: a 'Twin-beam' injection device and a field reversed mirror concept. The latter potentially offers fluxes approaching 10 16 n/cm 2 s but involves more speculative technology. (Auth.)

  20. The use of computer graphics to assist in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1995-01-01

    A number of complex inspection and repair tasks have been carried out inside Nuclear Electric's gas cooled reactors using multi-link manipulators. The paper describes how computer graphics are integrated into the manipulator system and perform a key role during many phases of a project. Graphics simulation is used to assist with conceptual design, detailed path planning, rehearsals, quality assurance, demonstrations and, critically, as an on-line tool to assist the real-time operation of the equipment whilst performing complex in-reactor tasks. The techniques used are illustrated by recent examples of use to assist the inspection and repair of in-reactor components. (Author)

  1. Technology assessment HTR. Part 3. Economics of new concept of the modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Lako, P.

    1996-06-01

    In this study the economic feasibility of new concepts of the High Temperature Reactor were investigated. These new concepts are characterized as inherently safe. The different concepts were used as industrial heat/power reactors and compared with a gas fired Steam and Gas turbine installation. The best economic advantages are offered by a HTR with a Thorium/Uranium cycle as compared with a gas fired steam- and gas turbine. 6 figs, 9 tabs, 21 refs

  2. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  3. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    International Nuclear Information System (INIS)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO 2 fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject

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

    International Nuclear Information System (INIS)

    Azpitarte, O.; Ramilo, L.

    2013-01-01

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

  5. Concept and designs of new-generation fast reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1993-01-01

    This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

  6. 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

  7. Supercritical-pressure, once-through cycle light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi

    2001-01-01

    The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

  8. Secret high-temperature reactor concept for inertial fusion

    International Nuclear Information System (INIS)

    Monsler, M.J.; Meier, W.R.

    1983-01-01

    The goal of our SCEPTRE project was to create an advanced second-generation inertial fusion reactor that offers the potential for either of the following: (1) generating electricity at 50% efficiency, (2) providing high temperature heat (850 0 C) for hydrogen production, or (3) producing fissile fuel for light-water reactors. We have found that these applications are conceptually feasible with a reactor that is intrinsically free of the hazards of catastrophic fire or tritium release

  9. Emergency medical assistance programs for nuclear power reactors

    International Nuclear Information System (INIS)

    Linnemann, R.E.; Mettler, F.A. Jr.

    1977-01-01

    This paper deals with a simple but practical medical support of geographically distributed nuclear reactors in isolated areas. A staff of experts at a centre devote their full attention to accident prevention and preparedness at reactor sites. They establish and maintain emergency medical programs at reactor sites and nearby support hospitals. The emphasis is on first aid and emergency treatment by medical attendants who are not and cannot be experts in radiation but do know how to treat patients. (author)

  10. Contributions to safety studies for new concepts of nuclear reactors

    International Nuclear Information System (INIS)

    Perdu, F.

    2003-12-01

    The complete study of molten salt reactors, designed for a massive and durable nuclear energy production, must include neutronics, hydraulics and thermal effects. This coupled study, using the MCNP and Trio U codes, is undertaken in the case of the MSRE (molten salt reactor experiment) prototype. The obtained results fit very well the experiment. Their extrapolation suggests ways of improving the safety coefficients of power molten salt reactors. A second part is devoted to accelerator driven subcritical reactors, developed to incinerate radioactive waste.We propose a method to measure the prompt reactivity from the decay following a neutron pulse. It relies only on the distribution of times between generations, which is a characteristic of the reactor. This method is implemented on the results of the MUSE 4 experiment, and the obtained reactivity is accurate within 5%. (author)

  11. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  12. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    International Nuclear Information System (INIS)

    Mynatt, Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-01-01

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs

  13. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  14. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  15. The ultimate safe (US) Reactor: A concept for the third millenium

    International Nuclear Information System (INIS)

    Gat, U.

    1986-01-01

    The Ultimate Safe (U.S.) Reactor is based on a novel safety concept. Fission products in the reactor are allowed to accumulate only to a level at which they would constitute a harmless source term. Removal of fission products also removes the decay heat - the driving force for the source term. The reactor has no excess criticality and is controlled by the reactivity temperature coefficient. Safety is inherent and passive. Waste is removed from the site promptly

  16. Space nuclear reactor concepts for avoidance of a single point failure

    International Nuclear Information System (INIS)

    El-Genk, M. S.

    2007-01-01

    This paper presents three space nuclear reactor concepts for future exploration missions requiring electrical power of 10's to 100's kW, for 7-10 years. These concepts avoid a single point failure in reactor cooling; and they could be used with a host of energy conversion technologies. The first is lithium or sodium heat pipes cooled reactor. The heat pipes operate at a fraction of their prevailing capillary or sonic limit. Thus, when a number of heat pipes fail, those in the adjacent modules remove their heat load, maintaining reactor core adequately cooled. The second is a reactor with a circulating liquid metal coolant. The reactor core is divided into six identical sectors, each with a separate energy conversion loop. The sectors in the reactor core are neurotically coupled, but hydraulically decoupled. Thus, when a sector experiences a loss of coolant, the fission power generated in it will be removed by the circulating coolant in the adjacent sectors. In this case, however, the reactor fission power would have to decrease to avoid exceeding the design temperature limits in the sector with a failed loop. These two reactor concepts are used with energy conversion technologies, such as advanced Thermoelectric (TE), Free Piston Stirling Engines (FPSE), and Alkali Metal Thermal-to- Electric Conversion (AMTEC). Gas cooled reactors are a better choice to use with Closed Brayton Cycle engines, such as the third reactor concept to be presented in the paper. It has a sectored core that is cooled with a binary mixture of He-Xe (40 gm/mole). Each of the three sectors in the reactor has its own CBC and neutronically, but not hydraulically, coupled to the other sectors

  17. Self-consistent Analysis of a Blanket and Shielding of a Fusion Reactor Concept

    International Nuclear Information System (INIS)

    Kim, Suk Kwon; Hong, B. G.; Lee, D. W.; Kim, D. H.; Lee, Y. O.

    2008-01-01

    To develop the concept of a DEMO reactor, a tokamak reactor system analysis code has been developed at KAERI (Korea Atomic Energy Research Institute). The system analysis code incorporates prospects of the development of plasma physics and the technologies in a simple mathematical model and it helps to develop the concept of a fusion reactor and to identify the necessary R and D areas for a realization of the concept. In the system code, a plant power balance equation and a plasma power balance equation are solved to find plant parameters which satisfy the plasma physics and technology constraints, simultaneously. The outcome of the system analysis is to identify which areas of plasma physics and technologies and to what extent they should be developed for a realization of given fusion reactor concepts

  18. Heat-pipe liquid-pool-blanket concept for the Tandem Mirror Reactor

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Johnson, G.L.

    1981-01-01

    The blanket concept for the tandem mirror reactor described in this paper was developed to produce the medium temperature heat (approx. 850 to 950 K) for the General Atomic sulfur-iodine thermochemical process for producing hydrogen. This medium temperature heat from the blanket constitutes about 81% of the total power output of the fusion reactor

  19. The passive response of the Integral Fast Reactor concept to the chilled inlet accident

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1990-01-01

    Simple methods are described for bounding the passive response of a metal fueled liquid-metal cooled reactor to the chilled inlet accident. Calculation of these bounds for a prototype of the Integral Fast Reactor concept shows that failure limits --- eutectic melting, sodium boiling and fuel pin failure --- are not exceeded. 2 refs., 1 fig., 2 tabs

  20. 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

  1. Assessment of the slowly-imploding liner (LINUS) fusion reactor concept

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1980-01-01

    Prospects for the slowly-imploding liner (LINUS) fusion reactor concept are reviewed. The concept envisages the nondestructive, repetitive and reversible implosion of a liquid-metal cylindrical annulus (liner) onto field-reversed DT plasmoids. Adiabatic heating of the plasmoid to ignition at ultra-high magnetic fields results in a compact, high power density fusion reactor with unique solutions to several technological problems and potentially favorable economics

  2. Study of the reactor relevance of the NET design concept

    International Nuclear Information System (INIS)

    Reynolds, P.; Worraker, W.J.

    1987-08-01

    The objective of the study was to explore the reactor relevance of NET, i.e. whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration power reactor (DEMO). The main areas of study were those near to the plasma, namely the divertor, first wall and tritium breeding blanket. Other aspects which were investigated were tritium permeation and recovery, reactor maintenance, afterheat and effects of disruptions. The principal results of the study are briefly presented; the details of the work are given in fourteen appendices. These appendices were selected for INIS and indexed separately. The overall conclusion of the study is that the NET design is only partly relevant to the design requirements of a DEMO reactor. (U.K.)

  3. Some basic concepts of fast breeder reactor safeguards

    International Nuclear Information System (INIS)

    Tkharev, E.; Walford, F.J.

    1987-04-01

    The range of discussion topics of this report is restricted to a few key areas of safeguards importance at Fast Breeder Reactors (FBR) only. The differences between thermal and fast reactors that may have safeguards significance in the case of FBRs are listed. The FBR principles of design are mentioned. The relevant safeguards objectives and criteria are given. The fundamental issues for safeguarding FBR are treated. An outline safeguards approach is presented. Model inspection activities are mentioned. 4 figs

  4. Study on dual plant concept for the next generation boiling water reactors

    International Nuclear Information System (INIS)

    Sato, Takashi; Oikawa, Hirohide

    1999-01-01

    The paper presents the study results on the basic concept of dual BWRs. For the convenience, we call the concept here as Trial Study on BWR dual concept (TSBWR dual). The concept is general and applicable to all BWRs which have internal recirculation pumps (RIP). The TSBWR dual is a plant concept of dual BWRs contained in a same secondary containment building. The plant output is from 2 x l,350 MWe up to 2 x 1,700 MWe. This concept is mainly aiming at safety improvement and cost savings of the next generation BWRs. The TSBWR dual has two RPVs and two dry wells (DW). It has, however, only one wet well (WW) and only one R/B. The WW and the R/B are shared by the dual reactors. The operating floor is also shared by the two reactors. The TSBWR dual has both passive safety systems and active safety systems. They are also shared between the two reactors. A lot of sharing between the dual reactors enables significant cost savings accompanied by the power increase up to 3,400 MWe. Although the TSBWR dual consists of two reactors, the simplified cylindrical configuration of the key structures and reduction of the R/B height can minimize the plant construction period. The TSBWR dual provides a concept with which we can challenge to construct a dual BWR plant in the near future. (author)

  5. Heterotopic pregnancy in an assisted reproduction conception; case ...

    African Journals Online (AJOL)

    Infertility management by assisted reproduction techniques has had rapid increase. While there is robust evidence supporting the efficacy and safety of assisted reproduction technique (ART), complications are encountered. Heterotopic pregnancy, defined as the presence of both an intrauterine and an ectopic gestation, ...

  6. Interaction design concepts for a mobile personal assistant

    NARCIS (Netherlands)

    Nagata, S.F.; Oostendorp, H. van; Neerincx, M.A.

    2004-01-01

    The Personal Assistant for onLine Services (PALS) project aims to develop an intelligent interface that facilitates efficient user interaction through personalization and context awareness with commerce web sites on a handheld device. The types of assistance services and interaction support

  7. Concepts for space nuclear multi-mode reactors

    International Nuclear Information System (INIS)

    Myrabo, L.; Botts, T.E.; Powell, J.R.

    1983-01-01

    A number of nuclear multi-mode reactor power plants are conceptualized for use with solid core, fixed particle bed and rotating particle bed reactors. Multi-mode systems generate high peak electrical power in the open cycle mode, with MHD generator or turbogenerator converters and cryogenically stored coolants. Low level stationkeeping power and auxiliary reactor cooling (i.e., for the removal of reactor afterheat) are provided in a closed cycle mode. Depending on reactor design, heat transfer to the low power converters can be accomplished by heat pipes, liquid metal coolants or high pressure gas coolants. Candidate low power conversion cycles include Brayton turbogenerator, Rankine turbogenerator, thermoelectric and thermionic approaches. A methodology is suggested for estimating the system mass of multi-mode nuclear power plants as a function of peak electric power level and required mission run time. The masses of closed cycle nuclear and open cycle chemical power systems are briefly examined to identify the regime of superiority for nuclear multi-mode systems. Key research and technology issues for such power plants are also identified

  8. A preliminary concept of stochastic model of the tritium cycle in a fusion reactor

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1988-01-01

    A preliminary concept of stochastic model of the tritium circulation in a fusion reactor was elaborated in purpose of determining the necessary minimum and current tritium inventory in real circumstances. A random character of reactor operation was assumed what is especially valid in the starting phase being of particularly low reliability of the assembly. A system of differential equations with random initial conditions describing the tritium cycle was solved for both operation and break states of the reactor. The distribution of the moments and of the number of breaks in the reactor operation was discussed and the possibilities of further development of the present model are indicated. 5 refs., 2 figs. (author)

  9. Concept of object-oriented intelligent support for nuclear reactor designing

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gofuku, A.

    1991-01-01

    A concept of object-oriented intelligent CAD/CAE environment is proposed for the conceptual designing of advanced nuclear reactor system. It is composed of (i) object-oriented frame-structure database which represents the hierarchical relationship of the composite elements of reactor core and the physical properties, and (ii) object-oriented modularization of the elementary calculation processes, which are needed for reactor core design analysis. As an example practise, an object-oriented frame structure is constructed for representing a 3D configuration of a special fuel element of a space reactor design, by using a general-purpose expert system shell ESHELL/X. (author)

  10. Conception of electron beam-driven subcritical molten salt ultimate safety reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Alekseev, P.N.; Ignat`ev, V.V. [Kurchatov Institute, Moscow (Russian Federation)] [and others

    1995-10-01

    This paper is a preliminary sketch of a conception to develop the {open_quotes}ultimate safety reactor{close_quotes} using modern reactor and accelerator technologies. This approach would not require a long-range R&D program. The ultimate safety reactor could produce heat and electric energy, expand the production of fuel, or be used for the transmutation of long-lived wastes. The use of the combined double molten salt reactor system allows adequate neutron multiplication to permit using an electron accelerator for the initial neutron flux. The general parameters of such a system are discussed in this paper.

  11. Changing concepts of geologic structure and the problem of siting nuclear reactors: examples from Washington State

    International Nuclear Information System (INIS)

    Tabor, R.W.

    1986-01-01

    The conflict between regulation and healthy evolution of geological science has contributed to the difficulties of siting nuclear reactors. On the Columbia Plateau in Washington, but for conservative design of the Hanford reactor facility, the recognition of the little-understood Olympic-Wallowa lineament as a major, possibly still active structural alignment might have jeopardized the acceptability of the site for nuclear reactors. On the Olympic Peninsula, evolving concepts of compressive structures and their possible recent activity and the current recognition of a subducting Juan de Fuca plate and its potential for generating great earthquakes - both concepts little-considered during initial site selection - may delay final acceptance of the Satsop site. Conflicts of this sort are inevitable but can be accommodated if they are anticipated in the reactor-licensing process. More important, society should be increasing its store of geologic knowledge now, during the current recess in nuclear reactor siting

  12. Plant dynamics analyses of fast reactor concept: RAPID-A without any control rod

    International Nuclear Information System (INIS)

    Kambe, Mitsuru

    1996-01-01

    Plant dynamics analyses of a fast reactor concept RAPID-A without any control rod have been demonstrated in case of reactor startup and sudden change of the primary flow rate. RAIP-A concept involves Lithium Expansion Module (LEM) for inherent reactivity feedback, Lithium Injection Module (LIM) for inherent ultimate shutdown and Lithium Release Module (LRM) for automated reactor startup. LEM consists of Quick-LEM and Slow-LEM. Slow-LEM provides with moderate reactivity addition as decreasing temperature. Quick-LEM assures quick negative reactivity feedback as increasing temperature. Plant dynamics analyses revealed that reactor power is nearly proportional to the primary flow rate even if the flow rate increases suddenly. Fully automated reactor startup from the subcritical condition has been attempted by inserting reactivity at a constant rate by LRM. Allowable rate of reactivity addition has been obtained in respect to Quick-LEM reactivity worth. (author)

  13. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Nascimento, Jamil Alves do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 %Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pinch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  14. An integral metallic-fueled and lead-cooled reactor concept for the 4th generation reactor

    International Nuclear Information System (INIS)

    Santos, A. dos; Nascimento, J.A. do

    2002-01-01

    An Integral Lead Reactor (ILR) concept is proposed for the 4th generation reactor to be used in the future. The ILR is loaded with metallic fuel and cooled by lead. It was evaluated in the 300-1500 MWe power range with the Japanese Fast Set 2 cross sections library. This set was tested against several fast benchmarks and the criticality uncertainty was found to be 0.51 % Δk. The reactor is started with U-Zr and changes to the U-TRU-Zr-RE fuel in a stepwise way. In the equilibrium cycle, the burnup reactivity is less than β eff for a core of the order of 300 MWe, pin diameter of 10.4 mm and a pin-pitch to diameter ratio of 1.308. The lead void reactivity is negative for reactor power less than 750 MWe. There is a need to improve the nuclear data for the major actinides. (author)

  15. Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

    International Nuclear Information System (INIS)

    PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

    2003-01-01

    The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

  16. An intelligent safety system concept for future CANDU reactors

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1980-01-01

    A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

  17. Analysis of a small Fast Sodium Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Gilberti, Mauricio, E-mail: mgilber@eletronuclear.gov.br [Eletrobrás Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil); Velasquez, Carlos E.; Vargas, Matheus L.; Martins, Felipe; Costa, Antonella L.; Veloso, Maria Auxiliadora F.; Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    This paper presents the analyses and initial results of a Small Fast Sodium Reactor (SFSR) simulated using MCNPX. The goal is to build a nuclear model and determine the main core neutronic parameters over the cycle. Neutronics parameters such as burnup neutronic behavior, depletion fuel composition, absorbing elements, core reactivity control and reactivity coefficients that affect the reactor cooled by sodium along its operation cycle have been analyzed. The parameters are evaluated in terms of the reactivity coefficients at different cycle stages. The results present a comparison and discussion of the differences found between the model developed and some information available in the literature for a similar project. (author)

  18. Subcritical enhanced safety molten-salt reactor concept

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Ignatiev, V.V.; Men'shikov, L.I.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.; Krasnykh, A.K.; Rudenko, V.T.; Somov, L.N.

    1995-01-01

    The nuclear power and its fuel cycle safety requirements can be met in the main by providing nuclear power with subcritical molten salt reactors (SMSR) - 'burner' with an external neutron source. The utilized molten salt fuel is the decisive advantage of the SMSR over other burners. Fissile and fertile nuclides in the burner are solved in a liquid salt in the form of fluorides. This composition acts simultaneously as: a) fuel, b) coolant, c) medium for chemical partitioning and reprocessing. The effective way of reducing the external source power consists in the cascade neutron multiplication in the system of coupled reactors with suppressed feedback between them. (author)

  19. New concepts for the recovery and isotopic separation of tritium in fusion reactors

    International Nuclear Information System (INIS)

    Dombra, A.H.; Holtslander, W.J.; Miller, A.I.; Canadian Fusion Fuels Technology Project, Toronto, Ontario)

    1986-01-01

    New concepts for the recovery of tritium from light water coolant of LiPb blankets, and high-pressure helium coolant of Li-ceramic blankets are introduced. Application of these concepts to fusion reactors is illustrated with conceptual system designs for the anticipated NET blanket requirements. (author)

  20. Substantiation of physical concepts of fast reactors in Russia: experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

    2008-07-01

    The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

  1. Packed-fluidized-bed blanket concept for a thorium-fueled commercial tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Chi, J.W.H.; Miller, J.W.; Karbowski, J.S.; Chapin, D.L.; Kelly, J.L.

    1980-09-01

    A preliminary design of a thorium blanket was carried out as a part of the Commercial Tokamak Hybrid Reactor (CTHR) study. A fixed fuel blanket concept was developed as the reference CTHR blanket with uranium carbide fuel and helium coolant. A fixed fuel blanket was initially evaluated for the thorium blanket study. Subsequently, a new type of hybrid blanket, a packed-fluidized bed (PFB), was conceived. The PFB blanket concept has a number of unique features that may solve some of the problems encountered in the design of tokamak hybrid reactor blankets. This report documents the thorium blanket study and describes the feasibility assessment of the PFB blanket concept

  2. Building a dynamic code to simulate new reactor concepts

    International Nuclear Information System (INIS)

    Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.

    2012-01-01

    Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.

  3. Conceptual design of laser fusion reactor, SENRI-I - 1. concept and system design

    International Nuclear Information System (INIS)

    Ido, S.; Naki, S.; Norimatsu, T.

    1981-01-01

    Design features of a laser fusion reactor concept SENRI-I and new concepts are reviewed and discussed. The unique feature is the utilization of a magnetic field to guide and control the inner liquid Li flow. Basic requirements and typical parameters used in the design are presented. Items to be discussed are constitution of the system, performance of liquid Li flow, neutronics, thermo-electric cycle, fuel cycle and new concepts

  4. Perinatal outcome of monochorionic and dichorionic twins after spontaneous and assisted conception: a retrospective cohort study.

    Science.gov (United States)

    Hack, Karien E A; Vereycken, Marijn E M S; Torrance, Helen L; Koopman-Esseboom, Corine; Derks, Jan B

    2018-06-01

    The aim of this study was to compare pregnancy outcomes in twin pregnancies after assisted conception and spontaneous conception, according to chorionicity. Retrospective cohort study of 1305 twin pregnancies between 1995 and 2015. All spontaneous (n = 731) and assisted conception conceived (n = 574) twin pregnancies with antenatal care and delivery in University Medical Center Utrecht, the Netherlands, a tertiary obstetric care center were studied according to chorionicity. Maternal age and incidence of nulliparity were higher among the assisted conception twins. Hypertensive disorders also appeared to be more frequent in assisted conception pregnancies, which could largely be explained by the higher proportion of elderly nulliparous women in this group. Spontaneously conceived twins were born earlier than twins after assisted conception, with subsequent lower birthweights and more admissions to a neonatal intensive care unit with increased neonatal morbidity. Monochorionic twins had worse pregnancy outcomes compared with dichorionic twins, irrespective of mode of conception; monochorionic twins conceived by assisted reproduction had more neonatal morbidity (mainly respiratory distress syndrome and necrotizing enterocolitis) and late neonatal deaths compared with spontaneously conceived monochorionic twins. Spontaneously conceived twins have worse pregnancy outcome compared with twins after assisted conception, probably due to a lower incidence of monochorionicity in the assisted conception group. The already increased perinatal risks in monochorionic twins are even higher in monochorionic twins conceived after infertility treatments compared with spontaneously conceived monochorionic twins, which warrants extra attention to these high-risk pregnancies. © 2018 The Authors. Acta Obstetricia et Gynecologica Scandinavica published by John Wiley & Sons Ltd on behalf of Nordic Federation of Societies of Obstetrics and Gynecology (NFOG).

  5. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  6. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  7. Experimental studies of tritium barrier concepts for fusion reactors

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  8. An international view of surgically assisted conception and surrogacy tourism.

    Science.gov (United States)

    Ahmad, Nehaluddin

    2011-01-01

    Modern medicine, specifically assisted reproductive technology (ART), has overtaken the law in many jurisdictions around the world. New technologies and practices open a Pandora's Box of ethical, religious, social and legal questions, and may present a variety of significant legal problems to the courts and legislators. Surrogate motherhood and pregnancy through ART have both attracted controversy. Some groups condemn ART and want it banned while its supporters acknowledge there is a need for legislative guidelines and regulations. A proposed statute, the Assisted Reproductive Technique Services Act, aimed at regulating reproductive technologies, including surrogacy arrangements, will be introduced in the Malaysian parliament, probably in 2012, and the Assisted Reproductive Technology (Regulation) Bill 2010 is already before the Indian parliament. This paper will discuss several of the potential socio-legal issues surrounding ART in the light of the complex situation, with a comparative analysis of the Malaysian, USA, UK and Indian positions.

  9. Requirements, needs, and concepts for a new broad-application test reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Fletcher, C.D.; Denison, A.B.; Liebenthal, J.L.

    1992-01-01

    For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need

  10. Application of the integrated blanket-coil concept (IBC) to fusion reactors

    International Nuclear Information System (INIS)

    Embrechts, M.J.; Steiner, D.; Mohanti, R.; Duggan, W.

    1987-01-01

    A novel concept is proposed for combining the blanket and coil functions of a fusion reactor into a single component and several unique applications to fusion reactor embodiments are identified. The proposed concept takes advantage of the fact that lithium is a good electrical conductor in addition to being a unique tritium-breeding material capable of energy recovery and transport at high temperatures. This concept, designated the ''integrated-blanket-coil (IBC) concept'' has the potential for: allowing fusion reactor embodiments which are easier to maintain; making fusion reactors more compact with an intrinsic ultra-high mass power density (net kW/sub E//metric tonne); and enhancing the tritium breeding potential for special coil applications such as ohmic heating and bean identation. By assuming a sandwich construction for the IBC walls (i.e., a layered combination of a thin wall of structural material, insulator and structural materials) the magnetohydrodynamic (MHD)-induced pressure drops and associated pressure stresses are modest and well below design limits. Possible unique applications of the IBC concept have been investigated and include the IBC concept applied to the poloidal field (PF) coils, toroidal field (TF) coils, divertor coils, ohmic heating (OH) coils, and identation coils for bean shaping

  11. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  12. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  13. Inherent safe design of advanced high temperature reactors - concepts for future nuclear power plants

    International Nuclear Information System (INIS)

    Hodzic, A.; Kugeler, K.

    1997-01-01

    This paper discusses the applicable solutions for a commercial size High Temperature Reactor (HTR) with inherent safety features. It describes the possible realization using an advanced concept which combines newly proposed design characteristics with some well known and proven HTR inherent safety features. The use of the HTR technology offers the conceivably best solution to meet the legal criteria, recently stated in Germany, for the future reactor generation. Both systems, block and pebble bed ,reactor, could be under certain design conditions self regulating in terms of core nuclear heat, mechanical stability and the environmental transfer. 23 refs., 7 figs

  14. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  15. Spherical torus (ST) concept and its reactor implications

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lazarus, E.A.; Miller, R.L.; Carreras, B.A.; Hogan, J.T.; Krakowski, R.A.; Seed, T.J.; Zubrin, R.M.; Schnurr, N.M.

    1986-01-01

    A brief description of the spherical torus design is given. The design concept includes resistive demountable toroidal field coils, poloidal divertor for impurity control, oscillating-field current maintenance, RF initiation and ramp-up of the plasma current, and flowing liquid-metal breeding blanket. 4 refs., 6 figs

  16. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  17. Concept and optimization of burning and transmutation reactor in nuclear fuel recycle system

    International Nuclear Information System (INIS)

    Marsodi; Mulyanto; Kitamoto, Asashi.

    1994-01-01

    Basic concept of B/T reactor, not only produces thermal energy but also performs burning and/or transmutation of MA and long-lived FPs, was introduced here based on numerical computation model. The advantage of nuclear reaction by thermal or fast neutron was combined conceptually with each other in order to maximize the overall B/T rate obtained by a composite system of fast and thermal reactor. According to the mass balance analysis of B/T reactors with P-T treatment, fast reactor hardened neutron energy may be effective for MA burning. Furthermore, a high flux reactor operated by fast or thermal neutron could be different from a reactor with high B/T rate or high capacity for loading of MA and/or long-lived FPs. The purpose of this study is to make clear the concept and the performance of fast and thermal B/T reactor designed under high neutron utilization for HLW disposal. (author)

  18. Gas Reactor International Cooperative Program. Interim report. Safety and licensing evaluaion of German Pebble Bed Reactor concepts

    International Nuclear Information System (INIS)

    1978-09-01

    The Pebble Bed Gas Cooled Reactor, as developed in the Federal Republic of Germany, was reviewed from a United States Safety and Licensing perspective. The primary concepts considered were the steam cycle electric generating pebble bed (HTR-K) and the process heat pebble bed (PNP), although generic consideration of the direct cycle gas turbine pebble bed (HHT) was included. The study examines potential U.S. licensing issues and offers some suggestions as to required development areas

  19. Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

    International Nuclear Information System (INIS)

    Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

    2001-01-01

    An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

  20. Students' conceptions of learning: using the ASSIST instrument ...

    African Journals Online (AJOL)

    The focus of the study has been on students' conceptions of learning in three South African Technikons amidst the changing circumstances of teaching and learning from subject-based to outcomes-based education. First-year students face a particular measure of unpreparedness as they graduate from a conventional high ...

  1. In-reactor testing of the closed cycle gas core reactor---the nuclear light bulb concept

    International Nuclear Information System (INIS)

    Gauntt, R.O.; Slutz, S.A.; Harms, G.A.; Latham, T.S.; Roman, W.C.; Rodgers, R.J.

    1993-01-01

    The Nuclear Light Bulb (NLB) concept is an advanced closed cycle space propulsion rocket engine design that offers unprecidented performance characteristics in terms of specific impulse (>1800 s) and thrust (>445 kN). The NLB is a gas-core nuclear reactor making use of thermal radiation from a high temperature U-plasma core to heat the hydrogen propellant to very high temperatures (∼4000 K). The following paper describes analyses performed in support of the design of in-reactor tests that are planned to be performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories in order to demonstrate the technical feasibility of this advanced concept. The tests will examine the stability of a hydrodynamically confined fissioning U-plasma under steady and transient conditions. Testing will also involve study of propellant heating by thermal radiation from the plasma and materials performance in the nuclear environment of the NLB. The analyses presented here include neutronic performance studies and U-plasma radiation heat-transport studies of small vortex-confined fissioning U-plasma experiments that are irradiated in the ACRR. These analyses indicate that high U-plasma temperatures (4000 to 9000 K) can be sustained in the ACRR for periods of time on the order of 5 to 20 s. These testing conditions are well suited to examine the stability and performance requirements necessary to demonstrate the feasibility of this concept

  2. Design concept of HYPER (HYbrid Power Extraction Reactor)

    International Nuclear Information System (INIS)

    Park, Won S.; Song, Tae Y.; Yu, Dong H.; Kim, Chang H.

    1999-01-01

    Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development called HYPER for the transmutation of nuclear waste and energy production through the transmutation process. Some major design features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Pb-Bi is adopted as a coolant and spallation target material. 1 GeV 13 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MWth power. The support ratio of HYPER for LWR units producing the same power is believed to be 5 to 6. (author)

  3. An integral reactor design concept for a nuclear co-generation plant

    International Nuclear Information System (INIS)

    Lee, D.J.; Kim, J.I.; Kim, K.K.; Chang, M.H.; Moon, K.S.

    1997-01-01

    An integral reactor concept for nuclear cogeneration plant is being developed at KAERI as an attempt to expand the peaceful utilization of well established commercial nuclear technology, and related industrial infrastructure such as desalination technology in Korea. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway to evaluate the characteristics of various passive safety concepts and provide the proper technical data for the conceptual design. This paper describes the preliminary safety and design concepts of the advanced integral reactor. Salient features of the design are hexagonal core geometry, once-through helical steam generator, self-pressurizer, and seismic resistant fine control CEDMS, passive residual heat removal system, steam injector driven passive containment cooling system. (author)

  4. Down syndrome screening in assisted conception twins: an iatrogenic medical challenge.

    Science.gov (United States)

    Ben-Ami, Ido; Maymon, Ron; Svirsky, Ran; Cuckle, Howard; Jauniaux, Eric

    2013-11-01

    The objective of this study was to provide a critical analysis of the impact of assisted conception on prenatal screening for Down syndrome (DS) in twin pregnancies and the value of various screening modalities for early detection of anomalies. The literature was searched using PubMed and the Cochrane Library focusing on prenatal screening and antenatal care of assisted-conception twin pregnancies. Serum screening alone is of limited value in detecting aneuploid twins, because the unaffected cotwin can "mask" the abnormal serum results of an affected one. In addition, this test can designate the pregnancy as at high risk but not identify the affected fetus. Nuchal translucency (NT) screening is the best available modality and a highly effective screening method for twin pregnancies. Among twins, NT alone has a 69% DS detection rate, first-trimester combined NT and serum biochemistry has a 72% DS detection rate, and an integrated screen will have an 80% DS detection rate at a 5% FPR. The data in the literature concerning the effect of assisted conception on maternal serum screening markers in twin pregnancies are scarce. Down syndrome screening in assisted-conception twins presents clinical and technical challenges. Therefore, assisted-conception twins need close monitoring from conception to delivery, by a practitioner familiar with the available screening modalities and their relative accuracy.

  5. Concept of an accelerator-driven subcritical research reactor within the TESLA accelerator installation

    International Nuclear Information System (INIS)

    Pesic, Milan; Neskovic, Nebojsa

    2006-01-01

    Study of a small accelerator-driven subcritical research reactor in the Vinca Institute of Nuclear Sciences was initiated in 1999. The idea was to extract a beam of medium-energy protons or deuterons from the TESLA accelerator installation, and to transport and inject it into the reactor. The reactor core was to be composed of the highly enriched uranium fuel elements. The reactor was designated as ADSRR-H. Since the use of this type of fuel elements was not recommended any more, the study of a small accelerator-driven subcritical research reactor employing the low-enriched uranium fuel elements began in 2004. The reactor was designated as ADSRR-L. We compare here the results of the initial computer simulations of ADSRR-H and ADSRR-L. The results have confirmed that our concept could be the basis for designing and construction of a low neutron flux model of the proposed accelerator-driven subcritical power reactor to be moderated and cooled by lead. Our objective is to study the physics and technologies necessary to design and construct ADSRR-L. The reactor would be used for development of nuclear techniques and technologies, and for basic and applied research in neutron physics, metrology, radiation protection and radiobiology

  6. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    Energy Technology Data Exchange (ETDEWEB)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

  7. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    International Nuclear Information System (INIS)

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc

  8. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Tsige-Tamirat, H.; Ammirabile, L.; D' Agata, E.; Fuetterer, M.; Ranguelova, V. [European Commission, Joint Research Centre, Institute for Energy, Westerduinweg 3, 1755LE Petten (Netherlands)

    2010-07-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  9. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ammirabile, L.; D'Agata, E.; Fuetterer, M.; Ranguelova, V.

    2010-01-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  10. Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

    International Nuclear Information System (INIS)

    K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

    2005-01-01

    Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

  11. Three core concepts for producing uranium-233 in commercial pressurized light water reactors for possible use in water-cooled breeder reactors

    International Nuclear Information System (INIS)

    Conley, G.H.; Cowell, G.K.; Detrick, C.A.; Kusenko, J.; Johnson, E.G.; Dunyak, J.; Flanery, B.K.; Shinko, M.S.; Giffen, R.H.; Rampolla, D.S.

    1979-12-01

    Selected prebreeder core concepts are described which could be backfit into a reference light water reactor similar to current commercial reactors, and produce uranium-233 for use in water-cooled breeder reactors. The prebreeder concepts were selected on the basis of minimizing fuel system development and reactor changes required to permit a backfit. The fuel assemblies for the prebreeder core concepts discussed would occupy the same space envelope as those in the reference core but contain a 19 by 19 array of fuel rods instead of the reference 17 by 17 array. An instrument well and 28 guide tubes for control rods have been allocated to each prebreeder fuel assembly in a pattern similar to that for the reference fuel assemblies. Backfit of these prebreeder concepts into the reference reactor would require changes only to the upper core support structure while providing flexibility for alternatives in the type of fuel used

  12. Innovation future reactors: the differential model of equivalent parameters. A decisional tool for conception

    International Nuclear Information System (INIS)

    Coudray, R.; Eid, M.; Fiorini, G.L.

    1992-01-01

    The study is developed in the frame of future reactors activities undertaken by the 'Commissariat a l'Energie Atomique' in France. The objective is to build the SAFETY and a QUALITY of the safety taking criteria, performances, experience... into account. In this purpose, we use the probabilistic reliability, maintainability, availability concepts, for the design and for the clustering of the parametric tendencies

  13. The integral fast reactor (IFR) concept: Physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  14. The integral fast reactor (IFR) concept: physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  15. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  16. Once-through cycle, supercritical-pressure light water cooled reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

    2001-07-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  17. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  18. Once-through cycle, supercritical-pressure light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.

    2001-01-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  19. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    International Nuclear Information System (INIS)

    Hughes, Joel T.; Blandford, Edward D.

    2016-01-01

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  20. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

    2016-07-15

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  1. 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

  2. Integral design concepts of advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-11-01

    Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessary in the small and medium range and need to be located closer to the load centres. The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas. The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development

  3. German concept and status of the disposal of spent fuel elements from German research reactors

    International Nuclear Information System (INIS)

    Komorowski, K.; Storch, S.; Thamm, G.

    1995-01-01

    Eight research reactors with a power ≥ 100 kW are currently being operated in the Federal Republic of Germany. These comprise three TRIGA-type reactors (power 100 kW to 250 kW), four swimming-pool reactors (power 1 MW to 10 MW) and one DIDO type reactor (power 23 MW). The German research reactors are used for neutron scattering for basic research in the field of solid state research, neutron metrology, for the fabrication of isotopes and for neutron activation analysis for medicine and biology, for investigating the influence of radiation on materials and for nuclear fuel behavior. It will be vital to continue current investigations in the future. Further operation of the German research reactors is therefore indispensable. Safe, regular disposal of the irradiated fuel elements arising now and in future operation is of primary importance. Furthermore, there are several plants with considerable quantities of spent fuel, the safe disposal of which is a matter of urgency. These include above all the VKTA facilities in Rossendorf and also the TRIGA reactors, where disposal will only be necessary upon decommissioning. The present paper report is concerned with the disposal of fuel from the German research reactors. It briefly deals with the situation in the USA since the end of 1988, describes interim solutions for current disposal requirements and then mainly concentrates on the German disposal concept currently being prepared. This concept initially envisages the long-term (25--50 years) dry interim storage of fuel elements in special containers in a central German interim store with subsequent direct final disposal without reprocessing of the irradiated fuel

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

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.

    2004-01-01

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

  5. The KALIMER-600 Reactor Core Design Concept with Varying Fuel Cladding Thickness

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Jang, Jin Wook; Kim, Yeong Il

    2006-01-01

    Recently, Korea Atomic Energy Research Institute (KAERI) has developed a 600MWe sodium cooled fast reactor, the KALIMER-600 reactor core concept using single enrichment fuel. This reactor core concept is characterized by the following design targets : 1) Breakeven breeding (or fissile-self-sufficient) without any blanket, 2) Small burnup reactivity swing ( 23 n/cm 2 ). In the previous design, the single enrichment fuel concept was achieved by using the special fuel assembly designs where non-fuel rods (i.e., ZrH 1.8 , B 4 C, and dummy rods) were used. In particular, the moderator rods (ZrH 1.8 ) were used to reduce the sodium void worth and the fuel Doppler coefficient. But it has been known that this hydride moderator possesses relatively poor irradiation behavior at high temperature. In this paper, a new core design concept for use of single enrichment fuel is described. In this concept, the power flattening is achieved by using the core region wise cladding thicknesses but all non-fuel rods are removed to simplify the fuel assembly design

  6. Evaluation of potential blanket concepts for a Demonstration Tokamak Hybrid Reactor

    International Nuclear Information System (INIS)

    Chapin, D.L.; Chi, J.W.H.; Kelly, J.L.

    1978-01-01

    An evaluation has been made of several different blanket concepts for use in a near-term Demonstration Tokamak Hybrid Reactor (DTHR), whose main objective would be to produce a significant amount of fissile fuel while demonstrating the feasibility of the tokamak hybrid reactor concept. The desirability of a simple design using proven technology plus a proliferation resistant fuel cycle led to the selection of a low temperature and pressure water-cooled, zircaloy clad ThO 2 blanket concept to breed 233 U. The nuclear performance and thermal-hydraulics characteristics of the blanket were evaluated to arrive at a consistent design. The blanket was found to be feasible for producing a significant amount of fissile fuel even with the limited operating conditions and blanket coverage in the DTHR

  7. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1978-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  8. Thermal hydraulic analyses of two fusion reactor first wall/blanket concepts

    International Nuclear Information System (INIS)

    Misra, B.; Maroni, V.A.

    1977-01-01

    A comparative study has been made of the thermal hydraulic performance of two liquid lithium blanket concepts for tokamak-type reactors. In one concept lithium is circulated through 60-cm deep cylindrical modules oriented so that the module axis is parallel to the reactor minor radius. In the other concept helium carrying channels oriented parallel to the first wall are used to cool a 60-cm thick stagnant lithium blanket. Paralleling studies were carried out wherein the thermal and structural properties of the construction materials were based on those projected for either solution-annealed 316-stainless steel or vanadium-base alloys. The effects of limitations on allowable peak structural temperature, material strength, thermal stress, coolant inlet temperature, and pumping power/thermal power ratio were evaluated. Consequences to thermal hydraulic performance resulting from the presence of or absence of a divertor were also investigated

  9. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    International Nuclear Information System (INIS)

    Klein, Andrew; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-01

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  10. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  11. Evaluation of an assisted convection system for integrated reactors

    International Nuclear Information System (INIS)

    Patruno, Luciano

    2006-01-01

    This work consists of the analyze of changing to an assisted convection system the CAREM-25 nuclear power plant design.The work is focused on the design, construction and measurement of an experimental hydraulic loop which contains a jet pump.There is also a characterization of parameters such as flow rate, hydrostatic pressure, efficiency, velocity profiles and development length using different tools such as theoretical formulations, numeric simulations and experimental data. New set points are found in CAREM thermo hydraulic parameters which take the thermal power to up to 150MW, and the water flow rate to 615 kg/s.To achieve this goals, the use of twelve jet pumps is proposed at a flow rate coefficient of 10.The results on the static pressure provided by the jet pump are calculated theoretically and compared with experimental data.There is an extrapolation of the results to higher flow rate values to be applied to the CAREM-25 design.Concerning the experimental work, the results show a good match on flow rate between the measured data and the theoretical results.There is a reasonable match in the coefficient of flow rate (M), even when the measurement uncertainties are quite high (about 13 %).It is also shown an efficiency measurement, which is up to 14 % [es

  12. SCW Pressure-Channel Nuclear Reactors: Some Design Features and Concepts

    International Nuclear Information System (INIS)

    Duffey, R.B.; Pioro, I.L.; Gabaraev, B.A.; Kuznetsov, Yu. N.

    2006-01-01

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950's and 1960's in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33 -- 35% to about 40 -- 45%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$ 1000 US/kW). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625 deg. C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia. Design features related to both channels and fuel bundles are discussed in this paper. Also, Russian experience with operating supercritical steam heaters at NPP is presented. The main conclusion is that development of SCW pressure-channel nuclear reactors is feasible and significant benefits can be expected over other thermal energy systems. (authors)

  13. Safety concept of high-temperature reactors based on the experience with AVR and THTR

    International Nuclear Information System (INIS)

    Wachholz, Winfried; Kroeger, Wolfgang

    1990-01-01

    In the Federal Republic of Germany a reactor is considered safe if verification has been furnished that the requirements contained in paragraph 7 of the Federal German Atomic Energy Act are met for this reactor: demonstration of sufficient precautions against damage required according to the actual state of the art, and especially compliance with the dose rate limits for normal operation and accidental conditions. These requirements result in a deterministic multi-stage safety concept with specified requirements for the engineered safety systems. In recent years, proposals for enhanced safety of nuclear power reactors or a radical change in safety philosophy have been made. This is characterised by 'inherently safe', 'super safe' and similar slogans. A quantitative definition of these requirements has not yet been established, but it is clear as a common objective that the event of beyond design basis accidents evacuation, relocation, and large scale contamination of ground should not occur. As a consequence of the Chernobyl accident the safety of all the NPPs in Germany has been reviewed. This analysis was completed for the THTR reactor in 1988. The same has been done for AVR reactor. The final evaluation of the HTR specific safety features have been fully confirmed. The HTR concepts under development are based on this experience. The HTR-Modul unit is currently being designed

  14. The Integral Fast Reactor concept: Today's hope for tomorrow's electrical energy needs

    International Nuclear Information System (INIS)

    Dwight, C.C.; Phipps, R.D.

    1989-01-01

    Acid rain and the greenhouse effect are getting more attention as their impacts on the environment become evident around the world. Substantial evidence indicates that fossil fuel combustion for electrical energy production activities is a key cause of those problems. A change in electrical energy production policy is essential to a stable, healthy environment. That change is inevitable, it's just a matter of when and at what cost. Vision now, instead of reaction later, both in technological development and public perception, will help to limit the costs of change. The Integral Fast Reactor (IFR) is a visionary concept developed by Argonne National Laboratory that involves electrical energy production through fissioning of heavy metals by fast neutrons in a reactor cooled by liquid sodium. Physical characteristics of the coolant and fuel give the reactor impressive characteristics of inherent and passive safety. Spent fuel is pyrochemically reprocessed and returned to the reactor in the IFR's closed fuel cycle. Advantages in waste management are realized, and the reactor has the potential for breeding, i.e., producing as much or more fuel than it uses. This paper describes the IFR concept and shows how it is today's hope for tomorrow's electrical energy needs. 14 refs., 1 fig., 1 tab

  15. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  16. Concept of an inherently-safe high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro

    2012-01-01

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  17. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  18. Report of the Office of Nuclear Reactor Regulation technical assistance task force

    International Nuclear Information System (INIS)

    1981-11-01

    In 1981, the Director of the Office of Nuclear Reactor Regulation (NRR) of the Nuclear Regulatory Commission (NRC) chartered a task force to assess the office program of technical assistance and to recommend improvements. The task force divided the technical assistance program into four areas, and the practices in each area were assessed through a series of surveys of staff, management, and contractor personnel. The task force placed emphasis in its interview and assessment process on the problem areas that exist in the technical assistance program. The report thus reflects a weight on the faults found as a result of the inquiries made. The four major areas of technical assistance contracting studied were program planning, program management and execution, program control and management information systems, and program administration and coordination

  19. Parents' contributions to the narrative identity of offspring of donor-assisted conception.

    Science.gov (United States)

    Kirkman, Maggie

    2003-12-01

    Donated sperm, eggs, and embryos are an increasing feature of assisted reproduction; people conceived in this way have different genetic and social histories. Although most offspring of donor-assisted conception are ignorant of their genetic history, recipient parents must negotiate increasing demand for full disclosure to offspring. This paper illustrates some of the reasons parents give for not telling their children, underlines the experience of many parents of being uncertain of how to go about telling, presents information from some parents who have endeavoured to be open with their children about conception from very early childhood, and discusses implications for the narrative identity of offspring of donor-assisted conception. Recipient parents (n=55) and offspring (12) from Australia; Canada, US, England, and Argentina were interviewed and subsequently consulted about the development of their narrative accounts and the way in which these have been interpreted. Parental narratives were found to be located along a continuum, broadly encompassing: (1). Parents who intend to exclude donor-assisted conception from the narratives they construct for their children, (2). parents who are uncertain about what they want to do, or confused about the best way to disclose and discuss donor conception with their children and (3). those have incorporated the donor in their children's narratives from the beginning. From interviews with offspring and on the basis of human rights issues and the increasing salience of genetic knowledge, it is concluded that disclosure to offspring before adolescence should be encouraged.

  20. Liquid metal versus gas cooled reactor concepts for a turbo electric powered space vehicle

    International Nuclear Information System (INIS)

    Carre, F.; Proust, E.; Schwartz, J.P.

    1985-01-01

    Recent CNES/CEA prospective studies of an orbit transfer vehicule to be launched by ARIANE V, emphasize the advantage of the Brayton cycle over the thermionics and thermoelectricity, in minimizing the total mass of 100 to 300 kWsub(e) power systems under the constraint specific to ARIANE of a radiator area limited to 95 m 2 . The review of candidate reactor concepts for this application, finally recommends both liquid metal and gas cooled reactors, for their satisfactory adaptation to a reference Brayton cycle and for the available experience from the terrestrial operation of comparable systems

  1. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  2. Temperature feedback effects in a supercritical water reactor concept with multiple heat-up steps

    Energy Technology Data Exchange (ETDEWEB)

    Barragan-Martinez, A.M., E-mail: albrm29@yahoo.com [Universidad Nacional Autonoma de Mexico, Departamento de Sistemas Energeticos, Facultad de Ingenieria, Jiutepec, Mor (Mexico); Espinosa-Paredes, G.; Vazquez-Rodriguez, A., E-mail: gepe@xanum.uam.mx, E-mail: vara@xanum.uam.mx [Universidad Autonoma Metropolitana-Iztapalapa, Area de Ingenieria en Rescursos Energeticos, Col. Vicentina (Mexico); Martin-del-Campo, C.; Francois, J.L., E-mail: cecilia.martin.del.campo@gmail.com, E-mail: juan.louis.francois@gmail.com [Universidad Nacional Autonoma de Mexico, Departamento de Sistemas Energeticos, Facultad de Ingenieria, Jiutepec, Mor (Mexico)

    2014-07-01

    The Supercritical Water Cooled Reactor (SCWR) is one of the most promising and innovative designs selected by the Generation IV International Forum. One of the concepts being studied is the High Performance Light Water Reactor (HPLWR), which is the European version of the SCWR. In this paper we present the numerical analysis of the behavior of a HPLWR with temperature feedback effects. The neutronic process, the heat transfer in the fuel rod and the thermalhydraulics in the core of the HPLWR were considered in this study. The neutronic calculations were performed with HELIOS-2 and the obtained results were used to evaluate the reactivity due to fuel temperature and supercritical water density. (author)

  3. Temperature feedback effects in a supercritical water reactor concept with multiple heat-up steps

    International Nuclear Information System (INIS)

    Barragan-Martinez, A.M.; Espinosa-Paredes, G.; Vazquez-Rodriguez, A.; Martin-del-Campo, C.; Francois, J.L.

    2014-01-01

    The Supercritical Water Cooled Reactor (SCWR) is one of the most promising and innovative designs selected by the Generation IV International Forum. One of the concepts being studied is the High Performance Light Water Reactor (HPLWR), which is the European version of the SCWR. In this paper we present the numerical analysis of the behavior of a HPLWR with temperature feedback effects. The neutronic process, the heat transfer in the fuel rod and the thermalhydraulics in the core of the HPLWR were considered in this study. The neutronic calculations were performed with HELIOS-2 and the obtained results were used to evaluate the reactivity due to fuel temperature and supercritical water density. (author)

  4. Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-11-01

    The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

  5. A HUMAN AUTOMATION INTERACTION CONCEPT FOR A SMALL MODULAR REACTOR CONTROL ROOM

    Energy Technology Data Exchange (ETDEWEB)

    Le Blanc, Katya; Spielman, Zach; Hill, Rachael

    2017-06-01

    Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Additionally, many new reactor concepts, both full scale and small modular reactors, are proposing increased automation and reduced staffing as part of their concept of operations. However, research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. There is a need to address the question of how to achieve high performance and efficiency of high levels of automation without degrading human performance. One example of a new NPP concept that will utilize greater degrees of automation is the SMR concept from NuScale Power. The NuScale Power design requires 12 modular units to be operated in one single control room, which leads to a need for higher degrees of automation in the control room. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed.

  6. Planning a new research reactor for AECL: The MAPLE-MTR concept

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

    1992-01-01

    AECL Research is assessing its needs and options for future irradiation research facilities. A planning team has been assembled to identify the irradiation requirements for AECL's research programs and compile options for satisfying the irradiation requirements. The planning team is formulating a set of criteria to evaluate the options and will recommend a plan for developing an appropriate research facility. Developing the MAPLE Materials Test Reactor (MAPLE-MTR) concept to satisfy AECL's irradiation requirements is one option under consideration by the planning team. AECL is undertaking this planning phase because the NRU reactor is 35 years old and many components are nearing the end of their design life. This reactor has been a versatile facility for proof testing CANDU components and fuel designs because the CANDU irradiation environment was simulated quite well. However, the CANDU design has matured and the irradiation requirements have changed. Future research programs will emphasize testing CANDU components near or beyond their design limits. To provide these irradiation conditions, the NRU reactor needs to be upgraded. Upgrading and refurbishing the NRU reactor is being considered, but the potentially large costs and regulatory uncertainties make this option very challenging. AECL is also developing the MAPLE-MTR concept as a potential replacement for the NRU reactor. The MAPLE-MTR concept starts from the recent MAPLE-X10 design and licensing experience and adapts this technology to satisfy the primary irradiation requirements of AECL's research programs. This approach should enable AECL to minimize the need for major advances in nuclear technology (e.g., fuel design, heat transfer). The preliminary considerations for developing the MAPLE-MTR concept are presented in this report. A summary of AECL's research programs is presented along with their irradiation requirements. This is followed by a description of safety criteria that need to be taken into

  7. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 degrees C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs

  8. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289 degrees C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  9. Environmentally assisted cracking in light water reactors. Semiannual report July 1996 - December 1996

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J.

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds

  10. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289 degree C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  11. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  12. A boiling-water reactor concept for low radiation exposure based on operating experience

    International Nuclear Information System (INIS)

    Koine, Y.; Uchida, S.; Izumiya, M.; Miki, M.

    1983-01-01

    A review of boiling-water reactor (BWR) operating experience indicates the significant role of water chemistry in determining the radiation dose rate contributing to occupational exposure. The major contributor among the radioactive species involved is identified as 60 Co, produced by neutron activation of 59 Co originating from structural materials. Iron crud, a fine solid form of corrosion product in the reactor water, is also shown to enhance the radiation dose rate. A theoretical study, supported by the operating experience and an extensive confirmatory test, led to the computerized analytical model called DR CRUD which is capable of predicting long-term radiation dose buildup. It accounts for the mechanism of radiation buildup through corrosion products such as irons, cobalts and other radioactive elements; their generation, transport, activation, interaction and deposition in the reactor coolant system are simulated. A scoping analysis, using this model as a tool, establishes the base line of the BWR concept for low occupational exposure. The base line consists of a set of target values for an annual exposure of 200 man.rem in an 1100 MW(e) BWR unit. They are the parameters that will be built into the design such as iron and cobalt inputs to the reactor water, and the capability of the reactor and the condensate purification system. Applicable means of technology are identified to meet the targets, ranging from improved water chemistry to the purification technique, optimized material selection and the recommended operational procedure. Extensive test programmes provide specifications of these means for use in BWRs. Combinations of their application are reviewed to define the concept of reduced exposure. Analytical study verifies the effectiveness of the proposed BWR concept in achieving a low radiation dose rate; occupational exposure is reduced to 200 man.rem/a. (author)

  13. Ambient Assisted Living Systems in the Context of Human Centric Sensing and IoT Concept

    DEFF Research Database (Denmark)

    Zaric, Nicola; Pejanovic-Djurisic, Milica; Mihovska, Albena Dimitrova

    2016-01-01

    This paper proposes the concept of Human Centric Sensing in the context of Internet of Things and Ambient Assisted Living. The paper uses a case study to present and analyze the proposed idea, and identifies the main challenges and open issues that require research and policy attention....

  14. The design features and safety concepts of the nuclear heating reactor developed in China

    International Nuclear Information System (INIS)

    Zheng Wenxiang; Wang Dazhong

    1995-01-01

    Based on the specific conditions of the nuclear heat applications and the development objectives of the advanced reactors, the nuclear heating reactor (NHR) exploited in China has adhered to the new safety concepts and been designed with a number of advanced features, including the integrated arrangement, full power natural circulation capacity, self-pressurized performance, dynamically-hydraulic control rod drive and passive safety systems, so that higher standard of safety as well as simplification in the plant systems and improvement in economic viability has been achieved. This paper describes the special consideration in the design as well as the main design features and safety concepts of the NHR. Some experimental and analytical results are also presented to demonstrate the NHR safety features

  15. Design concepts and status of the Korean next generation reactor (KNGR)

    International Nuclear Information System (INIS)

    Cho, Sung Jae; Kim, Han Gon

    1999-01-01

    The national project to develop KNGR, a 4000 MWth evolutionary advanced light water reactor (ALWR), has been organized in three phases according to the development status in 1992. During the first phase, the top-tier design requirements and the design concepts to meet the requirements had been established. The project is currently in the second phase of which the major objective is to complete the basic design sufficient to confirm the plant safety. This paper describes the overall design concepts and status of the KNGR briefly which developed and/or being developed through the project. (author)

  16. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  17. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  18. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  19. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  20. A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

    Science.gov (United States)

    Bloomfield, H. S.; Heller, J. A.

    1986-01-01

    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

  1. A feasibility assessment of nuclear reactor power system concepts for the NASA growth Space Station

    International Nuclear Information System (INIS)

    Bloomfield, H.S.; Heller, J.A.

    1986-01-01

    A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of space station related concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination

  2. A neutronic assessment of the new Spherical Cermets Fuel concept for the BWR-PB reactor

    International Nuclear Information System (INIS)

    Benchrif, A.; Chetaine, A.; Amsil, H.; Bounakhla, M.

    2010-01-01

    The tri-structural-isotopic (TRISO) fuel directly cooled by boiling light water is used in the boiling water reactor with pebble-bed coated particles (BWR-PB). At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating layer during a fuel cycle. So, the objectives of this paper, on the one hand is to evaluate some neutronic parameters of a new fuel concept, Spherical Cermets Fuel (SCF), for a BWR-PB reactor. On the other hand, to assess the fact of SCF fuel concept on the fuel assembly lifetime and the burn-up characteristic. All the parameters as well as Infinite Multiplication Factor, Spectrum Index, Instantaneous Conversion Ratio and Neutron Energy Spectrum was calculated then compared for the TRISO and the SCF fuel concept. It can be seen from the assessment of fuel assembly burn-up characteristics that the normalised neutron spectra of all the assembly's parts pointed out a thermal spectrum for the SCF fuel assembly's parts than the TRISO one. The SCF fuel element increase the assembly life time about 6.1 EFPY corresponding 8000 MWd/t. So, the fuel assembly can be operated for a reasonably long period without outside refuelling. The difference in the assembly lifetime might leads to SCF fuel concept adopted, because the geometry and concept of TRISO fuel particles are wholly different to SCF ones. (author)

  3. Concepts in developing technical means of accident shutdown of nuclear reactor

    International Nuclear Information System (INIS)

    Ionajtis, R.R.; Mikhajlov, M.P.; Cherkashov, Yu.M.

    1992-01-01

    Logic for realization of multistage (echelon) reactor accident shutdown system (ASS) is proposed on the basis of general safety concepts (OPB-88). ASS includes the basis stage with traditional composition of member systems (executive, control, providing ones), auxiliary (doubling) on the other principle of action and insuring (with direct action). Structural schemes of the system as a whole and member subsystems are presented. Recommendations on developing executive and control subsystems are given

  4. The basic concepts of a fuel-power detector for nuclear power reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.

    1979-01-01

    Fuel power is proposed as an alternative to neutron or gamma-ray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. (author)

  5. Concept of passive safe small reactor for distributed energy supply system

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Nakajima, Nobuya; Sawada, Ken-ichi; Yoritsune, Tsutomu; Shimada, Shoichiro; Nakano, Yoshihiro; Yonomoto, Taisuke; Takahashi, Hiroki

    2003-01-01

    This paper presents a concept of a Passive Safe Small Reactor for Distributed energy supply system (PSRD). The PSRD is an integrated-type PWR with reactor thermal power of 100 to 300 MW aimed at supplying electricity, district heating, etc. In design of the PSRD, high priority is laid on enhancement of safety as well as improvement of economy. Safety is enhanced by the following means: i) Extreme reduction of pipes penetrating the reactor vessel, by limiting to only those of the steam, the feed water and the safety valves, ii) Adoption of the water filled containment and the passive safety systems with fluid driven by natural circulation force, and iii) Adoption of the in-vessel type control rod drive mechanism, accompanying a passive reactor shut-down. To comply with a severe operation condition of PSRD, material of the ball bearing with graphite retainer has been selected by test. For improvement of economy, simplification of the reactor system and long operation of the core are achieved. Optimization of core design concerning the burnable poison ensures the burn-up of 28 GWd/t for low enriched UO 2 fuel rods. (author)

  6. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

  7. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  8. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2016-09-15

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  9. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    International Nuclear Information System (INIS)

    Garcia C, E. Y.; Francois L, J. L.

    2016-09-01

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  10. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289 degrees C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

  11. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    International Nuclear Information System (INIS)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β min is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β min , resulting in a list of candidate designs that possess the β value that is larger than the β min . The proposed methodology can also be applied to purposes other than technological foresight

  12. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

  13. Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

    International Nuclear Information System (INIS)

    Hejzlar, Pavel; Davis, Cliff B.

    2004-01-01

    A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners

  14. Effects of a Computer-Assisted Concept Mapping Learning Strategy on EFL College Students' English Reading Comprehension

    Science.gov (United States)

    Liu, Pei-Lin; Chen, Chiu-Jung; Chang, Yu-Ju

    2010-01-01

    The purpose of this research was to investigate the effects of a computer-assisted concept mapping learning strategy on EFL college learners' English reading comprehension. The research questions were: (1) what was the influence of the computer-assisted concept mapping learning strategy on different learners' English reading comprehension? (2) did…

  15. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289 degrees C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  16. Feasibility study of ultra-long life fast reactor core concept - 028

    International Nuclear Information System (INIS)

    Kim, T.K.; Taiwo, T.A.

    2010-01-01

    An ultra-long life core concept is proposed targeting capital and operational cost reductions and ultra-high discharge burnup in a fast reactor system. The core concept is achieved by de-rating the power density and adopting annular core geometry to maintain criticality for more than 40 years without refueling. The ultra-long life core has a specific power of ∼10 MW/t and an average driver fuel discharge burnup of ∼300 GWd/t. It is assumed such ultra-high burnup fuel can be developed within an advanced fuel cycle program. Several benefits are expected from the ultra-long life core concept such as capital and operational cost reductions, low proliferation risk, and effectively holding LWR spent fuel without disposal until technologies for a closed nuclear fuel cycle are developed and deployed. As future work, safety analysis, development of the advanced core cooling methods, and comparative cost analysis are expected. (authors)

  17. Post-partum depressive symptoms and medically assisted conception: a systematic review and meta-analysis.

    Science.gov (United States)

    Gressier, F; Letranchant, A; Cazas, O; Sutter-Dallay, A L; Falissard, B; Hardy, P

    2015-11-01

    Does medically assisted conception increase the risk of post-partum depressive symptoms? Our literature review and meta-analysis showed no increased risk of post-partum depressive symptoms in women after medically assisted conception. Women who conceive with medically assisted conception, which can be considered as a stressful life event, could face an increased risk of depressive symptoms. However, no previous meta-analysis has been performed on the association between medically assisted conception and post-partum depressive symptoms. A systematic review with electronic searches of PubMed, ISI Web of Knowledge and PsycINFO databases up to December 2014 was conducted to identify articles evaluating post-partum depressive symptoms in women who had benefited from medically assisted conception compared with those with a spontaneous pregnancy. Meta-analyses were also performed on clinically significant post-partum depressive symptoms according to PRISMA guidelines. From 569 references, 492 were excluded on title, 42 on abstract and 17 others on full-text. Therefore, 18 studies were included in the review and 8 in the meta-analysis (2451 women) on clinically significant post-partum depressive symptoms after medically assisted conception compared with a spontaneous pregnancy. A sensitivity meta-analysis on assisted reproductive technologies and spontaneous pregnancy (6 studies, 1773 women) was also performed. The quality of the studies included in the meta-analyses was evaluated using the Strengthening the Reporting of Observational Studies in Epidemiology Statement for observational research. The data were pooled using RevMan software by the Cochrane Collaboration. Heterogeneity between studies was assessed from the results of the χ(2) and I(2) statistics. Biases were assessed with funnel plots and Egger's test. A fixed effects model was used for the meta-analyses because of the low level of heterogeneity between the studies. The systematic review of studies examining

  18. Concept of a Prestressed Cast Iron Pressure Vessel for a Modular High Temperature Reactor

    International Nuclear Information System (INIS)

    Steinwarz, Wolfgang; Bounin, Dieter

    2014-01-01

    High Temperature Reactors (HTR) are representing one of the most interesting solutions for the upcoming generation of nuclear technology, especially with view to their inherent safety characteristics. To complete the safety concept of such plants already in the first phase of the technical development, Prestressed Cast Iron Pressure Vessels (PCIV) instead of the established forged steel reactor pressure vessels have been considered under the aspect of safety against bursting. A longterm research and development work, mainly performed in Germany, showed the excellent features of this technical solution. Diverse prototypic vessels were tested and officially proven. Design studies confirmed the feasibility of such a vessel concept also for Light Water Reactor types, too. The main concept elements of such a burst-proof vessel are: Strength and tightness functions are structurally separated. The tensile forces are carried by the prestressing systems consisting of a large number of independent wires. Compressive forces are applied to the vessel walls and heads. These are segmented into blocks of ductile cast iron. All cast iron blocks are prestressed to high levels of compression. The sealing function is assigned to a steel liner fixed to the cast iron blocks. The prestressing system is designed for an ultimate pressure of 2.3 times the design pressure. The prestress of the lids is designed for gapping at a much smaller pressure. Therefore, a drop of pressure will always occur before loss of strength (“leakage before failure”). In addition to these safety features further technical as well as economic aspects generate favorable assessment criteria: high design flexibility, feasibility of large vessel diameters; advantageous conditions for transport, assembly and decommissioning due to the segmented construction; advantage of workshop manufacturing; high-level quality control of components. Nowadays, considering the globally newly standardized safety requirements

  19. Lessons learned from the safety assistance program for soviet-designed reactors

    International Nuclear Information System (INIS)

    Steinberg, N.

    1999-01-01

    Two examples of nuclear power situation were compared in this conference paper - the situation in Lithuania and the situation in the Ukraine. Based on the examples mentioned, author conclude that the effectiveness of the Multi-National Safety Assistance Program for Soviet -Designed Reactors in a given recipient country does not depend, in practice, on engineering issues. The principal aspects that determine this effectiveness are: first, the level of safety culture in the country, beginning at the Governmental level but also at the level of the senior managers of nuclear power. The other important factor which contributes is the availability of a well-developed national program for upgrading NPP safety. The economical well-being of nuclear power and of the country as a whole also has a major effect on the effectiveness of the western technical assistance programs that are trying to upgrade reactor safety in a particular recipient country. And finally, international community should have well coordinated and well substantiated safety assistance program for specific country

  20. The concept and contribution of the Federal Government as regards the improvement of reactor safety in Middle and Eastern Europe

    International Nuclear Information System (INIS)

    Gast, K.; Breest, H.C.

    1992-01-01

    An increase in information on reactor safety in the Middle and Eastern European countries makes it clear that extensive assistance measures are urgently needed. This assistance can only be given on the basis of extensive international collaboration. A differenciated analysis and organization of the necessary assistance measures with specific reference to the countries and the facilities in question must be carried out. Within the framework of international assistance programmes, appropriate coordinating commissions in which banks with international experience are involved should be established. On the whole, the imperative increase in the safety of reactors in Middle and Eastern Europe will be a difficult and expensive task which can only be mastered if a joint effort is made within the framework of an international commission which addresses itself to the problem of reactor safety. (orig./HSCH) [de

  1. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  2. Possibility of a pressurized water reactor concept with highly inherent heat removal following capability

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Murao, Yoshio

    1995-01-01

    If the core power inherently follows change in heat removal rate from the primary coolant system within small thermal expansion of the coolant which can be absorbed in a practical size of pressurizer, reactor systems may have more safety and load following capability. In order to know possibility and necessary conditions of a concept on reactor core and primary coolant system of a pressurized water reactor (PWR) with such 'highly inherent heat removal following capability', transient analyses on an ordinary two-loop PWR have been performed for a transient due to 50% change in heat removal with the RETRAN-02 code. The possibility of a PWR concept with the highly inherent heat removal following capability has been demonstrated under the conditions of the absolute value of ratio of the coolant density reactivity coefficient to the Doppler reactivity coefficient more than 10x10 3 kg·cm 3 which is two to three times larger than that at beginning of cycle (BOC) in an ordinary PWR and realized by elimination of the chemical shim, the 12% lower average linear heat generation rate of 17.9 kW/m, and the 1.5 times larger pressurizer volume than those of the ordinary PWR. (author)

  3. Extraction of Water from Martian Regolith Simulant via Open Reactor Concept

    Science.gov (United States)

    Trunek, Andrew J.; Linne, Diane L.; Kleinhenz, Julie E.; Bauman, Steven W.

    2018-01-01

    To demonstrate proof of concept water extraction from simulated Martian regolith, an open reactor design is presented along with experimental results. The open reactor concept avoids sealing surfaces and complex moving parts. In an abrasive environment like the Martian surface, those reactor elements would be difficult to maintain and present a high probability of failure. A general lunar geotechnical simulant was modified by adding borax decahydrate (Na2B4O7·10H2O) (BDH) to mimic the 3 percent water content of hydrated salts in near surface soils on Mars. A rotating bucket wheel excavated the regolith from a source bin and deposited the material onto an inclined copper tray, which was fitted with heaters and a simple vibration system. The combination of vibration, tilt angle and heat was used to separate and expose as much regolith surface area as possible to liberate the water contained in the hydrated minerals, thereby increasing the efficiency of the system. The experiment was conducted in a vacuum system capable of maintaining a Martian like atmosphere. Evolved water vapor was directed to a condensing system using the ambient atmosphere as a sweep gas. The water vapor was condensed and measured. Processed simulant was captured in a collection bin and weighed in real time. The efficiency of the system was determined by comparing pre- and post-processing soil mass along with the volume of water captured.

  4. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  5. An innovative fuel design concept for improved light water reactor performance and safety. Final technical report

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1995-07-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. The purpose of this research was to explore a technique for extending fuel performance by thermally bonding LWR fuel with a non-alkaline liquid metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide (UO 2 ) fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Due to the thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high conductivity liquid metal thermally bonds the fuel to the cladding, and eliminates the large temperature change across the gap, while preserving the expansion and pellet loading capabilities. The resultant lower fuel temperature directly impacts fuel performance limit margins and also core transient performance. The application of liquid bonding techniques to LWR fuel was explored for the purposes of increasing LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) has been developed under the program to analyze the in-reactor performance of the liquid metal bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of Liquid Metal Bonded LWR fuel

  6. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  7. Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

    International Nuclear Information System (INIS)

    Yu, Zhang-cheng; Xie, Heng

    2014-01-01

    The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

  8. Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

    International Nuclear Information System (INIS)

    Popa-Simil, Liviu

    2008-01-01

    Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

  9. Concept of automated system for spent fuel utilization ('Reburning') from compact nuclear reactors

    International Nuclear Information System (INIS)

    Ianovski, V.V.; Lozhkin, O.V.; Nesterov, M.M.; Tarasov, N.A.; Uvarov, V.I.

    1997-01-01

    On the basic concept of an automated system of nuclear power installation safety is developed the utilization project of spent fuel from compact nuclear reactors. The main features of this project are: 1. design and creation of the mobile model-industrial installation; 2. development of the utilization and storage diagram of the spent fuel from compact nuclear reactors, with the specific recommendation for the natatorial means using both for the nuclear fuel reburning, for its transportation in places of the storage; 3. research of an opportunity during the utilization process to obtain additional power resources, ozone and others to increase of justifying expenses at the utilization; 4. creation of new generation engineering for the automation of remote control processes in the high radiation background conditions. 7 refs., 1 fig

  10. Techno-economic assessment of membrane assisted fluidized bed reactors for pure H_2 production with CO_2 capture

    International Nuclear Information System (INIS)

    Spallina, V.; Pandolfo, D.; Battistella, A.; Romano, M.C.; Van Sint Annaland, M.; Gallucci, F.

    2016-01-01

    Highlights: • Membrane reactors improve the overall efficiency of H_2 production up to 20%. • Respect to conventional reforming, the H_2 yield increases from 12% to 20%. • The COH is reduced of at least 220% using membrane reactors. • FBMR capture 72% of CO_2 with a specific cost of 8 eur/tonn_C_O_2_. • MA-CLR can reach 90% of CO_2 avoided with same cost of FTR. - Abstract: This paper addresses the techno-economic assessment of two membrane-based technologies for H_2 production from natural gas, fully integrated with CO_2 capture. In the first configuration, a fluidized bed membrane reactor (FBMR) is integrated in the H_2 plant: the natural gas reacts with steam in the catalytic bed and H_2 is simultaneously separated using Pd-based membranes, and the heat of reaction is provided to the system by feeding air as reactive sweep gas in part of the membranes and by burning part of the permeated H_2 (in order to avoid CO_2 emissions for heat supply). In the second system, named membrane assisted chemical looping reforming (MA-CLR), natural gas is converted in the fuel rector by reaction with steam and an oxygen carrier (chemical looping reforming), and the produced H_2 permeates through the membranes. The oxygen carrier is re-oxidized in a separate air reactor with air, which also provides the heat required for the endothermic reactions in the fuel reactor. The plants are optimized by varying the operating conditions of the reactors such as temperature, pressures (both at feed and permeate side), steam-to-carbon ratio and the heat recovery configuration. The plant design is carried out using Aspen Simulation, while the novel reactor concepts have been designed and their performance have been studied with a dedicated phenomenological model in Matlab. Both configurations have been designed and compared with reference technologies for H_2 production based on conventional fired tubular reforming (FTR) with and without CO_2 capture. The results of the analysis show

  11. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  12. Reactor physical program in the frame of the MSR-SPHINX transmuter concept development

    International Nuclear Information System (INIS)

    Hron, M.; Mikisek, M.

    2008-01-01

    In the frame of the R and D program for the Molten Salt Reactor (MSR) - SPHINX (Spent Hot fuel Incinerator by Neutron flux) concept, which has been under development in the Czech Republic as an actinide burner in resonance neutron spectrum and a radionuclide transmuter in a well-thermalized neutron spectrum, and namely its reactor physical part, the relatively broad experimental activities have been involved in the program, recently, which will serve for a validation of computer codes and verification of design inputs for designing of a demonstration unit of the MSR-type. The experimental program, which has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors, will be in the proposed next stage of the program focused on a large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0, which may allow to modify it to experimental zero power salt reactor SR-0. There has been a preparatory stage of the project called EROS started in the year 2006 and new experiments with MSR-type zones irradiated by cyclotron based neutron source are planned at the end of 2007 and should go on in the year 2008. There will be a brief description of the so far prepared and performed experimental programs introduced in the paper. (authors)

  13. Accident Tolerant Fuel Concepts for Light Water Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2016-06-01

    Nuclear fuel is a highly complex material that has been subject to continuous development over the past 40 years and has reached a stage where it can be safely and reliably irradiated up to 65 GWd/tU in commercial nuclear reactors. During this time, there have been many improvements to the original designs and materials used. However, the basic design of uranium oxide fuel pellets clad with zirconium alloy tubing has remained the fuel choice for the vast majority of commercial nuclear power plants. Severe accidents, such as those at the Three Mile Island and Fukushima Daiichi have shown that under such extreme conditions, nuclear fuel will fail and the high temperature reactions between zirconoi alloys and water will lead to the generation of hydrogen, with the potential for explosions to occur, daming the plant further. Recognizing that the current fuel designs are vulnerable to severe accident conditions, tehre is renewed interesst in alternative fuel designs that would be more resistant to fuel failure and hydrogen production. Such new fuel designs will need to be compatible with existing fuel and reactor systems if they are to be utilized in the current reactor fleet and in current new build designs, but there is also the possibility of new designs for new reactor systems. This publication provides a record of the Technical Meeting on Accident Tolerant Fuel Concepts for Light Water Reactors, held at Oak Ridge National Laboratories (ORNL), United States of America, 13-16 October 2014, to consider the early stages of research and development into accident tolerant fuel. There were 45 participants from 10 countries taking part in the meeting, with 32 papers organized into 7 sessions, of which 27 are included in this publication. This meeting is part of a wider investigation into such designs, and it is anticipated that further Technical Meetings and research programmes will be undertaken in this field

  14. Startup thaw concept for the SP-100 space reactor power system

    Science.gov (United States)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

    1990-01-01

    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

  15. Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

    1978-06-01

    Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

  16. Refrigeration requirements for fusion reactors based upon the theta-pinch concept

    International Nuclear Information System (INIS)

    Williamson, K.D. Jr.; King, C.R.

    1976-01-01

    Two refrigeration systems applicable to the theta-pinch fusion concept are described. The first is a 1100 W, 4.5 K refrigerator which will be used for testing superconducting NbTi Magnetic Energy Transfer and Storage (METS) coil systems. This unit is currently being installed and is to be operational by April 1977. The second unit is applicable to the Syllac Fusion Test Reactor (SFTR) and has been conceptually designed. This liquefier-refrigerator is about 22 times larger than those in existence at present and will require 12-MW input electrical power. It will provide 3045 kg/h of liquid helium at 4.5 K

  17. Comparative study for minor actinide transmutation in various fast reactor core concepts

    International Nuclear Information System (INIS)

    Ohki, S.

    2001-01-01

    A comparative evaluation of minor actinide (MA) transmutation property was performed for various fast reactor core concepts. The differences of MA transmutation property were classified by the variations of fuel type (oxide, nitride, metal), coolant type (sodium, lead, carbon dioxide) and design philosophy. Both nitride and metal fuels bring about 10% larger MA transmutation amount compared with oxide fuel. The MA transmutation amount is almost unchanged by the difference between sodium and lead coolants, while carbon dioxide causes a reduction by about 10% compared with those. The changes of MA transmutation property by fuel and coolant types are comparatively small. The effects caused by the difference of core design are rather significant. (author)

  18. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  19. Rapid and efficient proteolysis through laser-assisted immobilized enzyme reactors.

    Science.gov (United States)

    Zhang, Peng; Gao, Mingxia; Zhu, Shaochun; Lei, Jie; Zhang, Xiangmin

    2011-11-25

    In this report, laser radiation (808nm) for the first time was employed to enhance the efficiency of proteolysis through immobilized enzyme reactor (IMER). IMER based monolithic support was prepared in the fused-silica capillary via a simple two-step procedure including acryloylation on trypsin surface and in situ aqueous polymerization/immobilization. The feasibility and high efficiency of the laser-assisted IMER were demonstrated by the digestion of bovine serum albumin (BSA), cytochrome c (Cyt-c) and β-casein. The digestion process was achieved in 60s. The peptides were identified by MALDI-TOF-MS, yielding the sequence coverage of 33% for BSA, 73% for Cyt-c and 22% for β-casein. The comparisons between the in-solution digestion and on IMER reaction with/without laser assistance were made. To further confirm its efficiency in proteome analysis, the laser-assisted IMER was also applied to the analysis of one fraction of human serum sample through two-dimensional (2-D) separation of strong anion exchange/reversed-phase liquid chromatography (SAX/RPLC). After a database search, 49 unique peptides corresponding to 5 proteins were identified. The results showed that the laser-assisted IMER provides a promising platform for the high-throughput protein identification. Copyright © 2011 Elsevier B.V. All rights reserved.

  20. Accident tolerant high-pressure helium injection system concept for light water reactors

    International Nuclear Information System (INIS)

    Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

    2016-01-01

    Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

  1. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  2. On the concept of elasticity used in some fast reactor accident analysis codes

    International Nuclear Information System (INIS)

    Malmberg, T.

    1975-01-01

    The analysis to be presented will restrict attention to the elastic part of the elastic-plastic constitutive equation used in several Fast Reactor Accident Analysis Codes and originally applied by M.L. Wilkins: Calculation of Elastic-Plastic Flow, UCRL-7322, Rev. 1, Jan. 1969. It is shown that the used elasticity concept is within the frame of hypo-elasticity. On the basis of a test found by Bernstein it is proven that the state of stress is generally depending on the path of deformation. Therefore this concept of elasticity is not compatible with finite elasticity. For several simple deformation processes this special hypo-elastic constitutive equation is integrated to give a stress-strain relation. The path-dependence of this relation is demonstrated. Further the phenomenon of hypo-elastic yield under shear deformation is pointed out. The relevance to modelling material behaviour in primary containment analysis is discussed

  3. On the concept of elasticity used in some fast reactor accident analysis codes

    International Nuclear Information System (INIS)

    Malmberg, T.

    1975-01-01

    The analysis presented restricts attention to the elastic part of the elastic-plastic equation used in several Fast Reactor Accident Analysis Codes and originally applied by M.L. Wilkins: Calculation of Elastic-Plastic Flow, UCRL-7322, Rev. 1, Jan 1969. It is shown that the used elasticity concept is within the frame of hypo-elasticity. On the basis of a test found by Bernstein it is proven that the state of stress is generally depending on the path of deformation. Therefore this concept of elasticity is not compatible with finite elasticity. For several deformation processes this special hypo-elastic constitutive equation is integrated to give a stress-strain relation. The path-dependence of this relation is demonstrated. Further the phenomenon of hypo-elastic yield under shear deformation is pointed out. The relevance to modelling material behaviour in primary containment analysis is discussed. (Auth.)

  4. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  5. Computer network that assists in the planning, execution and evaluation of in-reactor experiments

    International Nuclear Information System (INIS)

    Bauer, T.H.; Froehle, P.H.; August, C.; Baldwin, R.D.; Johanson, E.W.; Kraimer, M.R.; Simms, R.; Klickman, A.E.

    1985-01-01

    For over 20 years complex, in-reactor experiments have been performed at Argonne National Laboratory (ANL) to investigate the performance of nuclear reactor fuel and to support the development of large computer codes that address questions of reactor safety in full-scale plants. Not only are computer codes an important end-product of the research, but computer analysis is also involved intimately at most stages of experiment planning, data reduction, and evaluation. For instance, many experiments are of sufficiently long duration or, if they are of brief duration, occur in such a purposeful sequence that need for speedy availability of on-line data is paramount. This is made possible most efficiently by computer assisted displays and evaluation. A purposeful linking of main-frame, mini, and micro computers has been effected over the past eight years which greatly enhances the speed with which experimental data are reduced to useful forms and applied to the relevant technological issues. This greater efficiency in data management led also to improvements in the planning and execution of subsequent experiments. Raw data from experiments performed at INEL is stored directly on disk and tape with the aid of minicomputers. Either during or shortly after an experiment, data may be transferred, via a direct link, to the Illinois offices of ANL where the data base is stored on a minicomputer system. This Idaho-to-Illinois link has both enhanced experiment performance and allowed rapid dissemination of results

  6. Conception of thermonuclear reactor with a shielding layer of the first wall

    International Nuclear Information System (INIS)

    Marin, S.V.

    1979-01-01

    Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

  7. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    Science.gov (United States)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  8. Improved structure and long-life blanket concepts for heliotron reactors

    International Nuclear Information System (INIS)

    Sagara, A.; Imagawa, S.; Mitarai, O.

    2005-01-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14m is selected to permit a blanket-shield thickness of about 1m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R and D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated. (author)

  9. Improved structure and long-life blanket concepts for heliotron reactors

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

  10. Recent developments in engineering and technology concepts for prospective tokamak fusion reactors

    International Nuclear Information System (INIS)

    Ford, G.W.K.

    1987-01-01

    The tokamak has become the most developed magnetic fusion system and it appears likely that break-even and possibly ignition will first be demonstrated in existing machines of this type. Yet larger tokamaks could also demonstrate the essential technologies for the production of useful power. World-wide, well over a hundred tritium-breeder/heat-removal blanket concepts have been devised and preliminary engineering design studies undertaken, but the effort deployed on breeding and power recovery systems has been very small compared with that assigned to plasma research and development. The European Communities' NET (Next European Torus) project may offer an opportunity to redress this imbalance. The NET pre-design stage now in progress for some three years has selected many of the best features of plasma and nuclear design from the world's total efforts in these fields, and the NET concept is described in this paper as exemplifying where magnetic fusion power reactor technology stands today. It is concluded that although there are numerous more advanced types of magnetic confinement fusion reactor at early stages of their physics development, the tokamak offers the best opportunity for the early demonstration of fusion power

  11. Use of the local-global concept in detecting component vibration in reactors

    International Nuclear Information System (INIS)

    Al-Ammar, M.A.

    1981-01-01

    The local-global concept, based on the detector adjoint function, was used to develop the response of a detector to an absorber vibrating in one dimension. A one-dimensional two-group diffusion code was developed to calculate the frequency dependent detector response as a function of detector and absorber positions for the coupled-core UTR-10 reactor. Results from this code indicated the best possible detector and absorber locations, where more detailed calculations were made using a two-group, three-dimensional diffusion code with finite detector and absorber volumes. An experiment was then designed, for the chosen positions, using a vibrating cadmium absorber with a detector on each side. The assembly was placed in the vertical central stringer of the reactor. Investigations were carried out for vibrations in two flux gradients and experimental data were analyzed in the frequency domain using a microcomputer-based data acquisition system. The experimental investigation showed the validity of the local-global concept. A normalized outputs cross power spectral density was developed that correctly predicted the different flux tilts in the two flux gradients. It was also shown that the frequency response of the local component had a wide plateau region. Monitoring the behavior of the normalized cross power spectral density was thought to be a promising indicator for the detection and localization of malfunctioning vibrating components. It might also be used to detect flux irregularities in the vicinity of a vibrating component

  12. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  13. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  14. Diagnostic mirror concept development for use in the complex environment of a fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krimmer, Andreas Joachim

    2016-07-01

    Light-based diagnostic systems of fusion reactors require optical mirrors to channel light through the structures surrounding the plasma. With increasing plasma volume, power and plasma burn time, the environmental conditions grow more demanding and new requirements arise. In this dissertation, the design of optical mirrors inside the vacuum chamber of the prototype reactor ITER (Latin ''the way'') and future fusion power plants are investigated. Comparing the state of the art with the boundary conditions close to the fusion plasma, existing mirror designs and choices for the reflective surface are evaluated. For the design, it is not the individual boundary conditions that are critical, but rather, their combination and the resulting interactions. Drawing from the existing designs, possible realizations for central functionality are discussed. Included in the discussion are substrate choice, mounting, adjustment and thermal contacting as well as positioning of the mirror assembly compatible with hot cell maintenance. Building on the general discussion, mirror concepts for the charge exchange recombination spectroscopy (CXRS) diagnostic system for the ITER plasma core are proposed and simulated. In addition, prototypes are manufactured and tested to assess critical aspects of the proposed design. Testing includes positioning by pins, manufacturing of a stainless steel substrate with fluid channels adapted to the mirror shape, and tests with an SiO{sub 2} /TiO{sub 2} dielectric coating under selected ITER conditions. As a result of the work, the fusion reactor mirror design considerations given in the principal design discussion can be used as a basis for other diagnostic systems as well. In the case of the core CXRS mirror concept for ITER, the basic suitability was shown and critical topics were identified where additional work is necessary.

  15. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  16. Diagnostic mirror concept development for use in the complex environment of a fusion reactor

    International Nuclear Information System (INIS)

    Krimmer, Andreas Joachim

    2016-01-01

    Light-based diagnostic systems of fusion reactors require optical mirrors to channel light through the structures surrounding the plasma. With increasing plasma volume, power and plasma burn time, the environmental conditions grow more demanding and new requirements arise. In this dissertation, the design of optical mirrors inside the vacuum chamber of the prototype reactor ITER (Latin ''the way'') and future fusion power plants are investigated. Comparing the state of the art with the boundary conditions close to the fusion plasma, existing mirror designs and choices for the reflective surface are evaluated. For the design, it is not the individual boundary conditions that are critical, but rather, their combination and the resulting interactions. Drawing from the existing designs, possible realizations for central functionality are discussed. Included in the discussion are substrate choice, mounting, adjustment and thermal contacting as well as positioning of the mirror assembly compatible with hot cell maintenance. Building on the general discussion, mirror concepts for the charge exchange recombination spectroscopy (CXRS) diagnostic system for the ITER plasma core are proposed and simulated. In addition, prototypes are manufactured and tested to assess critical aspects of the proposed design. Testing includes positioning by pins, manufacturing of a stainless steel substrate with fluid channels adapted to the mirror shape, and tests with an SiO_2 /TiO_2 dielectric coating under selected ITER conditions. As a result of the work, the fusion reactor mirror design considerations given in the principal design discussion can be used as a basis for other diagnostic systems as well. In the case of the core CXRS mirror concept for ITER, the basic suitability was shown and critical topics were identified where additional work is necessary.

  17. Concept of an immersive assistance system with augmented reality for the support of manual activities in radioactive production environments

    International Nuclear Information System (INIS)

    Eursch, Andreas A.

    2010-01-01

    The thesis on an immersive assistance system concept with augmented reality for the support of manual activities in radioactive production environments covers the following topics: analysis of the situation: production and use of radioactive materials, problem analysis of the work in the production facilities, necessity of manual activities, automation, prediction in hot cells; status of research and development; assistance system concept, immersive camera system; augmented reality support in hot cells; economic evaluation and generalization.

  18. Amster: a molten-salt reactor concept generating its own 233U and incinerating transuranium elements

    International Nuclear Information System (INIS)

    Lecarpentier, D.; Garzenne, C.; Vergnes, J.; Mouney, H.; Delpech, M.

    2002-01-01

    In the coming century, sustainable development of atomic energy will require the development of new types of reactors able to exceed the limits of the existing reactor types, be it in terms of optimum use of natural fuel resources, reduction in the production of long-lived radioactive waste, or economic competitiveness. Of the various candidates with the potential to meet these needs, molten-salt reactors are particularly attractive, in the light of the benefits they offer, arising from two fundamental features: - A liquid fuel does away with the constraints inherent in solid fuel, leading to a drastic simplification of the fuel cycle, in particular making in possible to carry out on-line pyrochemical reprocessing; - Thorium cycle and thermal spectrum breeding. The MSBR concept proposed by ORNL in the 1970's thus gave a breeding factor of 1.06, with a doubling time of about 25 years. However, given the tight neutron balance of the thorium cycle (the η of 233 U is about 2.3), MSBR performance is only possible if there are strict constraints set on the in-line reprocessing unit: all the 233 Pa must be removed from the core so that it can decay on the 233 U in no more than about ten days (or at least 15 tonnes of salt to be extracted from the core daily), and the absorbing fission products, in particular the rare earths, must be extracted in about fifty days. With the AMSTER MSR concept, which we initially developed for incinerating transuranium elements, we looked to reduce the mass of salt to be reprocessed in order to minimise the size and complexity of the reprocessing unit coupled to the reactor, and the quantity of transuranium elements sent for disposal, as this is directly proportional to the mass of salt reprocessed for extraction of the fission products. Given that breeding was not an absolute necessity, because the reactor can be started by incinerating the transuranium elements from the spent fuel assemblies of current reactors, or if necessary by loading

  19. Application of the Defense-in-Depth Concept in the Projects of New-Generation NPPs Equipped with VVER Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shvyryaev, Yu. V.; Morozov, V. B.; Kuchumov, A.Yu., E-mail: morozov@aep.ru [JSC Atomenergoproekt, Moscow (Russian Federation)

    2014-10-15

    The projects of new-generation NPPs equipped with VVER reactors are developed as projects the safety level of which is superior to that of NPPs that are currently in operation. The main design solutions adopted for implementing the defence-in-depth (DiD) concept in the projects of new-generation NPPs equipped with VVER reactors are briefly characterized in the paper. (author)

  20. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  1. Optimization of the deep-burner-modular helium reactor (DB-MHR) concept for actinide incineration

    International Nuclear Information System (INIS)

    Trakas, Ch.; Bruna, G.B.

    2005-01-01

    The paper summarizes studies performed on the General Atomics Deep-Burner Modular Helium Reactor (DB-MHR) concept-design carried out by Framatome ANP, Areva's joint subsidiary with Siemens. Feasibility and sensitivity studies as well as fuel-cycle studies with probabilistic methodology are presented. Emphasis is put on most attractive physical and computational aspects of the concept. Current investigations on design uncertainties, the future search for ways to improve the transmutation value in a double-stratum strategy, and the computational tools improvement are also presented. The Areva HTR, ANTARES, uses a similar prismatic core design. In that context, we revisited and optimized the Deep Burn concept. Typical values of transmutation ratio were established at 70%. Accounting for reactivity control needs estimated at 300 pcm, the cycle length can be estimated at 480 full-power days, giving an overall fuel irradiation time (6 fuel cycles) of about 10 calendar year, assuming as usual an average capacity factor of the plant of 80%. This study indicates a quite significant (96%) achievement for fissile material incineration. After a 500 year cooling-time, radiotoxicity (without fission products) is reduced by a factor 4

  2. The history and perspective of Romania-USA cooperation in the field of technologic transfer of TRIGA reactor concept

    International Nuclear Information System (INIS)

    Ciocaanescu, M.; Ionescu, M.

    1996-01-01

    The cooperation between Romania and the USA in the field of technologic transfer of nuclear research reactor technology began with the steady state 14 MW t TRIGA reactor, installed at INR Pitesti, Romania. It is the first in the range of TRIGA reactors proposed as a materials testing reactor. The first criticality was reached in November 19, 1979 and first operation at 14 MW t level was in February 1980. The paper will present the short history of this cooperation and the perspective for a new cooperation for building a Nuclear Heating Plant using the TRIGA reactor concept for demonstration purpose. The energy crisis is a world-wide problem which affects each country in different ways because the resources and the consumption are unfairly distributed. World-wide research points out that the fossil fuel sources are not to be considered the main energy sources for the long term as they are limited

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

    International Nuclear Information System (INIS)

    Obara, Toru; Sekimoto, Hiroshi

    1997-01-01

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

  4. Concept of the LORELEI Test Device for LOCA Experiment in the JHR Reactor

    International Nuclear Information System (INIS)

    Moran, N.; Ferry, L.; Azulay, A.; Mileguir, O.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis. Former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. In JHR material testing reactor, which is currently under construction, one significant experimental device is the LORELEI testing device. The objective is to examine the LOCA sequence influence on: thermo-mechanical behavior of the fuel clad, possible fuel relocation, corrosion at high temperature, oxidation, hydriding and resulted clad embrittment. The device is a single rod closed loop system placed on a displacement device inside a defined channel in the reflector. Several operational constrains on the device, as required by the reactor operational philosophy resulted quite a few challenges in the design. Constrains as: pre experimental re-irradiation phase under thermo-syphonic flow, application of active insulation to simulate the surrounding fuel, application of tensile force during refolding simulation, controlling the experiment with non-direct temperature measurement, etc. requires sophisticated solutions. The main objective of the conceptual design was to remove the uncertainties of those challenging requirements. The current presentation describes the approach applied defining the concept of the device, using sophisticated design combined with computational and experimental tools

  5. PRISM [Power Reactor Inherently Safe Module] design concept enhances waste management

    International Nuclear Information System (INIS)

    Thompson, M.L.; Berglund, R.C.

    1989-01-01

    PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs

  6. Nonthermal fusion reactor concept based on Hall-effect magnetohydrodynamics plasma theory

    International Nuclear Information System (INIS)

    Witalis, E.A.

    1988-01-01

    The failure of magnetic confinement controlled thermonuclear fusion research to achieve its goal is attributed to its foundation on the incomplete MHD plasma description instead of the more general HMHD (Hall-effect magnetohydrodynamics) theory. The latter allows for a certain magnetic plasma self-confinement under described stringent conditions. A reactor concept based on the formation, acceleration, and forced disintegration of magnetized whirl structures, plasmoids, is proposed. The four conventional MHD theory objections, i.e., absence of dynamo action, fast decay caused by resistivity, non-existence of magnetic self-confinement, and negligible non-thermal fusion yield, are shown not to apply. Support for the scheme from dense plasma focus research is pointed out. (orig.) [de

  7. Induced structural radioactivity inventory analysis of the base case aqueous ATW reactor concept

    International Nuclear Information System (INIS)

    Bezdecny, J.A.; Henderson, D.L.; Sailor, W.C.

    1993-01-01

    The purpose of the Los Alamos National Laboratory Accelerator Transmutation of Nuclear Waste (ATW) project is the substantial reduction in volume of this country's long-lived high-level radioactive waste in a safe and energy efficient manner. An evaluation of the Accelerator Transmutation of Nuclear Waste concept has four aspects; material balance, energy balance, performance and cost. An evaluation of the material balance compares the amount of long-lived high-level waste transmuted with the amount and type of waste created in the process. One component of the material balance is the activation of structural materials over the lifetime of the transmutation reactor. An activation analysis has been performed on four structure regions of the reaction vessel: the tungsten target; the lead target and annulus; the Zircalloy and aluminum tubing carrying the actinide slurry and; the stainless steel tank

  8. Application of the ALAP concept to occupational exposure at operating light water reactors

    International Nuclear Information System (INIS)

    Dickson, H.W.; Cottrell, W.D.; Jacobs, D.G.

    1975-01-01

    The application of the as-low-as-practicable (ALAP) concept to radiation exposure of workers at light-water reactors (LWR's) has recently received increased attention. The purpose of the project described is to investigate the means by which occupational exposure at operating LWR's can be reduced to the lowest practicable levels. Nine LWR stations, including 16 operating reactors, were studied in Phase I of the project to identify significant sources of exposure and to determine the magnitude of the exposures. A complete site review consists of compiling information from safety analysis reports, plant technical specifications, and radiation exposure records coupled with an on-site visit for discussions with plant personnel, observation of procedures, and measurement of radiation levels. In Phase II, specific problem areas are being studied in-depth with regard to corrective measures to reduce exposure. Information has been collected on exposure from valve maintenance and repair. Corrective measures will be evaluated with respect to ease of application and cost effectiveness. The results presented will serve as technical backup for the preparation of regulatory guides

  9. Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

    Directory of Open Access Journals (Sweden)

    Dalin Zhang

    2012-06-01

    Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

  10. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-06-15

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant.

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

    International Nuclear Information System (INIS)

    Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L.

    1991-08-01

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs

  12. High temperature reactor module power plant. Plant and safety concept June 1986 - 38.07126.2

    International Nuclear Information System (INIS)

    1986-06-01

    The modular HTR power plant is a universally applicable energy source for the co-generation of electricity, process steam or district heating. The modular HTR concept is characterized by the fact that standardized reactor units with power ratings of 200 MJ/s (so-called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR) are also available at higher power plant ratings. The safety features, the technical design and the mode of operation are briefly described in the following, taking a power plant with two HTR-Modules for the co-generation of electricity and process steam as an example. Due to its universal applicability and excellent safety features, the modular HTR power plant is suitable for erection on any site, but particularly on sites near other industrial plants or in densely populated areas. The co-generation of electricity and process steam or district heating with a modular HTR power plant as described here is primarily tailored to the requirements of industrial and communal consumers. The site for such a plant is a typical industrial one. The anticipated features of such sites were taken into consideration in the design of the modular HTR power plant

  13. Use of dwell time concept in fission product inventory assessment for CANDU reactors

    International Nuclear Information System (INIS)

    Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

    2003-01-01

    A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

  14. An optimized power conversion system concept of the integral, inherently-safe light water reactor

    International Nuclear Information System (INIS)

    Memmott, Matthew J.; Wilding, Paul R.; Petrovic, Bojan

    2017-01-01

    Highlights: • Three power conversion systems (PCS) for the I"2S-LWR are presented. • An optimization analyses was performed to evaluate these PCS alternatives. • The ideal PCS consists of 5 turbines, and obtains an overall efficiency of 35.7%. - Abstract: The integral, inherently safe light water reactor (I"2S-LWR) has been developed to significantly enhance passive safety capabilities while maintaining cost competitiveness relative to the current light water reactor (LWR) fleet. The compact heat exchangers of the I"2S-LWR preclude boiling of the secondary fluid, which decreases the probability of heat exchanger failure, but this requires the addition of a flash drum, which negatively affects the overall plant thermodynamic efficiency. A state of the art Rankine cycle is proposed for the I"2S-LWR to increase the thermodynamic efficiency by utilizing a flash drum with optimized operational parameters. In presenting this option for power conversion in the I"2S-LWR power plant, the key metric used in rating the performance is the overall net thermodynamic efficiency of the cycle. In evaluating the flash-Rankine cycle, three basic industrial concepts are evaluated, one without an intermediate pressure turbine, one with an intermediate turbine and one reheat stream, and one with an intermediate turbine and two reheat streams. For each configuration, a single-path multi-variable optimization is undertaken to maximize the thermal efficiency. The third configuration with an intermediate turbine and 2 reheat streams is the most effective concept, with an optimized efficiency of 35.7%.

  15. The Text of the Instrument Concerning the Agency's Assistance to Argentina for the Establishment of a Training Reactor Project

    International Nuclear Information System (INIS)

    1971-01-01

    The text of the Agreement between the Agency and the Governments of Argentina and the Federal Republic of Germany concerning the Agency's assistance for the establishment of a training reactor project in Argentina is reproduced herein for the information of all Members. The Agreement entered into force on 13 March 1970.

  16. The Text of the Instrument Concerning the Agency's Assistance to Argentina for the Establishment of a Training Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1971-02-23

    The text of the Agreement between the Agency and the Governments of Argentina and the Federal Republic of Germany concerning the Agency's assistance for the establishment of a training reactor project in Argentina is reproduced herein for the information of all Members. The Agreement entered into force on 13 March 1970.

  17. Environmentally assisted cracking in light water reactors - annual report, January-December 2001

    International Nuclear Information System (INIS)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2003-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to ∼2 x 10 21 n · cm -2 (E > 1 MeV) (∼3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at ∼325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several

  18. Laccase/mediator assisted degradation of triarylmethane dyes in a continuous membrane reactor.

    Science.gov (United States)

    Chhabra, Meenu; Mishra, Saroj; Sreekrishnan, Trichur Ramaswamy

    2009-08-10

    Laccase/mediator systems are important bioremediation agents as the rates of reactions can be enhanced in the presence of the mediators. The decolorization mechanism of two triarylmethane dyes, namely, Basic Green 4 and Acid Violet 17 is reported using Cyathus bulleri laccase. Basic Green 4 was decolorized through N-demethylation by laccase alone, while in mediator assisted reactions, dye breakdown was initiated from oxidation of carbinol form of the dye. Benzaldehyde and N,N-dimethyl aniline were the major end products. With Acid Violet 17, laccase carried out N-deethylation and in mediator assisted reactions, oxidation of the carbinol form of the dye occurred resulting in formation of formyl benzene sulfonic acid, carboxy benzene sulfonic acid and benzene sulfonic acid. Toxicity analysis revealed that Basic Green 4 was toxic and treatment with laccase/mediators resulted in 80-100% detoxification. The treatment of the textile dye solution using laccase and 2,2'-azino-di-(-ethylbenzothiazoline-6-sulfonic acid) (ABTS) was demonstrated in an enzyme membrane reactor. At a hydraulic retention time of 6h, the process was operated for a period of 15 days with nearly 95% decolorization, 10% reduction in flux and 70% recovery of active ABTS.

  19. Concepts of mental capacity for patients requesting assisted suicide: a qualitative analysis of expert evidence presented to the Commission on Assisted Dying.

    Science.gov (United States)

    Price, Annabel; McCormack, Ruaidhri; Wiseman, Theresa; Hotopf, Matthew

    2014-04-22

    In May 2013 a new Assisted Dying Bill was tabled in the House of Lords and is currently scheduled for a second reading in May 2014. The Bill was informed by the report of the Commission on Assisted Dying which itself was informed by evidence presented by invited experts.This study aims to explore how the experts presenting evidence to the Commission on Assisted Dying conceptualised mental capacity for patients requesting assisted suicide and examine these concepts particularly in relation to the principles of the Mental Capacity Act 2005. This study was a secondary qualitative analysis of 36 transcripts of oral evidence and 12 pieces of written evidence submitted by invited experts to the Commission on Assisted Dying using a framework approach. There was agreement on the importance of mental capacity as a central safeguard in proposed assisted dying legislation. Concepts of mental capacity, however, were inconsistent. There was a tendency towards a conceptual and clinical shift toward a presumption of incapacity. This appeared to be based on the belief that assisted suicide should only be open to those with a high degree of mental capacity to make the decision.The 'boundaries' around the definition of mental capacity appeared to be on a continuum between a circumscribed legal 'cognitive' definition of capacity (in which most applicants would be found to have capacity unless significantly cognitively impaired) and a more inclusive definition which would take into account wider concepts such as autonomy, rationality, voluntariness and decision specific factors such as motivation for decision making. Ideas presented to the Commission on Assisted Dying about mental capacity as it relates to assisted suicide were inconsistent and in a number of cases at variance with the principles of the Mental Capacity Act 2005. Further work needs to be done to establish a consensus as to what constitutes capacity for this decision and whether current legal frameworks are able to

  20. Concepts for the interim storage of spent fuel elements from research reactors in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Niephaus, D.; Bensch, D.; Quaassdorff, P.; Plaetzer, S.

    1997-01-01

    Research reactors have been operated in the Federal Republic of Germany since the late fifties. These are Material Test Reactors (MTR) and training, Research and Isotope Facilities of General Atomic (TRIGA). A total of seven research reactors, i.e. three TRIGA and four MTR facilities were still in operation at the beginning of 1996. Provisions to apply to the back-end of the fuel cycle are required for their continued operation and for already decommissioned plants. This was ensured until the end of the eighties by the reprocessing of spent fuel elements abroad. In view of impeding uncertainties in connection with waste management through reprocessing abroad, the development of a national back-end fuel cycle concept was commissioned by the Federal Minister of Education, Science, Research and Technology in early 1990. Development work was oriented along the lines of the disposal concept for irradiated light-water reactor fuel elements from nuclear power plants. Analogously, the fuel elements from research reactors are to be interim-stored on a long-term basis in adequately designed transport and storage casks and then be directly finally disposed without reprocessing after up to forty years of interim storage. As a first step in the development of a concept for interim storage, several sites with nuclear infrastructure were examined and assessed with respect to their suitability for interim storage. A reasonably feasible reference concept for storing the research reactor fuel elements in CASTOR MTR 2 transport and storage casks at the Ahaus interim storage facility (BZA) was evaluated and the hot cell facility and AVR store of Forschungszentrum Juelich (KFA) were proposed as an optional contingency concept for casks that cannot be repaired at Ahaus. Development work was continued with detailed studies on these two conceptual variants and the results are presented in this paper. (author)

  1. ATBR - a thorium breeder reactor concept for an early induction of thorium

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, Usha; Karthikeyan, R.; Ganesan, S.; Jain, R.P.; Kamat, S.U.

    1999-07-01

    A new reactor concept has been proposed for induction of thorium in an enriched uranium reactor. The neutronic characteristics of the fissile and fertile materials, have been exploited to arrive at optimal fuel assembly and core configurations. Each fuel assembly consists of an enriched uranium seed zone and a thoria blanket zone. They are in the form, of ring type fuel clusters. The fuel is contained in vertical pressure tubes placed in a hexagonal lattice array in D 2 O moderator. Boiling H 2 O coolant is used. The 235 U enrichment is about 5%. The thoria rods contain the 233 U bred in situ by irradiation of one batch load of mere thoria clusters (without the seed zone) for one fuel cycle in the same reactor. There is no need for external feed enrichment in thoria rods. Additionally some moveable thoria clusters are used for the purpose of xenon over-ride. The fissile production rate from the fertile material and the consumption rate of fissile inventory is judiciously balanced by the choice of U/Th fuel rod dia, the number and location of thoria rods in the fuel assembly and in the core. During steady state operation at rated power level, there in no need for any conventional control maneuvers such as change in soluble boron concentration or control rod movement as a function of burnup. Burnable poison rods are also not required. A very small reactivity fluctuation of ±2 mk in 300 effective full power days of operation is achieved and can be nearly met by coolant inlet enthalpy changes or moveable thoria clusters. Control is required only for cold shutdown of the reactor. The uranium as well as thoria rods achieve a fairly high burnup of 30-35 GWD/T at the time of discharge. Since the excess reactivity for hot full power operation is nearly zero at all time during the fuel cycle and since the coefficients of reactivity due to temperature and density variations of coolant are nearly zero by design, there is hardly any possibility of severe accidents involving

  2. Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Stosic, Zoran V. [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)], E-mail: Zoran.Stosic@areva.com; Brettschuh, Werner; Stoll, Uwe [AREVA NP GmbH, Koldestr. 16, 91052 Erlangen (Germany)

    2008-08-15

    reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact. A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.

  3. CATHARE simulation of transients for the 2400 MW gas fast reactor concept

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Messie, Anne; Geffraye, Genevieve; Malo, Jean-Yves; Bertrand, Frederic; Plancq, David

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is one of the six reactor concepts selected in the framework of the Generation IV forum and a high priority in the French Commissariat a l'Energie Atomique (CEA) R and D program on the Future Nuclear Energy Systems. A first design of this GFR2400 reactor has been completed by the CEA at the end of year 2005. The main characteristics of the concept are a 2400MW core based on plate type fuel elements, with an inlet temperature of 400degC and an outlet temperature of 850degC. The power conversion system is based on an indirect combined cycle with helium on the primary circuit, a Brayton cycle with a mixture of nitrogen and helium on the secondary circuit and a steam cycle on the tertiary circuit. In accidental situations, the use of the gas coolant circulation as the main way to remove the decay heat has been selected. A specific system (DHR system) has been designed: it consists of three loops (3 * 100% redundancy) in extension of the pressure vessel, equipped with heat exchangers and blowers. Between 2006 and 2007 a pre-conceptual study has been achieve, leading to the CEA milestone project of the 'GFR viability' at the end of year 2007. In the frame of this milestone, a wide range of CATHARE2 transients has been achieved to consolidate and improve the decay heat removal strategy; in particular the DHR blowers working on a large pressure range and the use of natural convection as a second way to remove decay heat. The paper first presents the CATHARE2 code applied to gas cooled reactor, focusing on the dedicated features included in the standard option of the code in order to obtain a multi-fluid reliable and performing tool. Then the modeling of the GFR2400 is presented, including the core, the vessel, the primary and secondary circuit with the turbo-machine, and a simplified tertiary circuit with boundary conditions. The decay heat removal loops (DHR loop) are also modeled, with a first circuit in helium and a secondary circuit in

  4. The use of LBB concept in French fast reactors: Application to SPX plant

    International Nuclear Information System (INIS)

    Turbat, A.; Deschanels, H.; Sperandio, M.

    1997-01-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented

  5. The use of LBB concept in French fast reactors: Application to SPX plant

    Energy Technology Data Exchange (ETDEWEB)

    Turbat, A.; Deschanels, H.; Sperandio, M. [and others

    1997-04-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented.

  6. Development of a double containment concept for the European pressurized water reactor

    International Nuclear Information System (INIS)

    Costaz, J.L.; Bonhomme, N.; L'Huby, Y.; Sidaner, J.F.

    1994-01-01

    This paper addresses the development of a double containment concept for the European Pressurized Water Reactor. Specification of containment leak tightness during severe hazards resulting from core melt scenarios is part of the safety goals defined for the EPR project. These safety goals include retention of molten core, mitigation of hydrogen deflagration or explosion risks and decay heat removal. The main new containment structural design loads which have been defined, including containment pressure and temperature conditions following possible postulated-core melt events, are recalled in the paper. The feasibility of a double containment with a prestressed concrete inner containment taking into account these new design loads but based upon experience gained within the well tested concept of concrete double wall containment used in 1400 MW nuclear power plants which have already been built in France, is presented. The main characteristics of such a prestressed inner containment are described. Limits and further possible optimization for even more severe design loads (including liner option) are indicated. Experimental works including a large scale mock up are already under way. (author). 2 refs., 4 figs

  7. Fast flux fluid fuel reactor: A concept for the next generation of nuclear power production

    International Nuclear Information System (INIS)

    Palmiotti, G.; Feldman, E.E.

    1999-01-01

    Nuclear energy has not become the preferred method of electrical energy production largely because of economic, safety, and proliferation concerns and challenges posed by nuclear waste disposal. Economies is the most important factor. To reduce the capital costs, the authors propose a compact configuration with a very high power density and correspondingly reduced reactor component sizes. Enhanced efficiency made possible by higher operating temperatures will also improve the economics of the design, and design simplicity will keep capital, operational, and maintenance costs down. The most direct solution to the nuclear waste problem is to eliminate waste production or, at least, minimize its amount and long-term radiotoxicity. This can be achieved by very high burnups, ideally 100%, and by the eventual transmutation of the long-lived fission products in situ. Very high burnups also improve the economics by optimal exploitation of the fuel. Safety concerns can be addressed by an inherently safe reactor design. Because of the intrinsic nature of nuclear materials, there probably is no definitive answer to proliferation concerns for systems that generate neutrons; however, it is important to minimize proliferation risks. The thorium cycle is a promising option because (a) plutonium is produced only in very small quantities, (b) the presence of 232 U makes handling the fuel very difficult and therefore proliferation resistant, and (c) 233 U is a fissile isotope that is less suitable than 239 Pu for making weapons and can be diluted with other uranium isotopes. An additional benefit of the thorium cycle is that it increases nuclear fuel resources by one order of magnitude. A fast flux fluid fuel reactor is a concept that can satisfy all the foregoing requirements. The fluid fuel systems have a very simple structure. Because integrity of the fuel is not an issue, these systems can operate at very high temperatures, can have high power densities, and can achieve very

  8. Experimental demonstration of the reverse flow catalytic membrane reactor concept for energy efficient syngas production. Part 2: Model development

    NARCIS (Netherlands)

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

    2007-01-01

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

  9. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  10. Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

    International Nuclear Information System (INIS)

    Ghyym, Seong H.

    1998-01-01

    In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

  11. Portable microwave assisted extraction: An original concept for green analytical chemistry.

    Science.gov (United States)

    Perino, Sandrine; Petitcolas, Emmanuel; de la Guardia, Miguel; Chemat, Farid

    2013-11-08

    This paper describes a portable microwave assisted extraction apparatus (PMAE) for extraction of bioactive compounds especially essential oils and aromas directly in a crop or in a forest. The developed procedure, based on the concept of green analytical chemistry, is appropriate to obtain direct in-field information about the level of essential oils in natural samples and to illustrate green chemical lesson and research. The efficiency of this experiment was validated for the extraction of essential oil of rosemary directly in a crop and allows obtaining a quantitative information on the content of essential oil, which was similar to that obtained by conventional methods in the laboratory. Copyright © 2013 Elsevier B.V. All rights reserved.

  12. An improved reactor system for small-scale fuel processing - the Optiformer concept

    Energy Technology Data Exchange (ETDEWEB)

    Silversand, Fredrik; Karlsson, Charlotte (Catator AB, Lund (Sweden))

    2008-08-15

    Catator AB (CAT) has evaluated a revised design of the previously described Ultraformer concept. In comparison to the original Ultraformer, the new reformer concept (Optiformer) shows enhanced performances with respect to thermo-mechanic durability and conversion efficiencies. The unit is also easy to manufacture and ordinary high-temperature steel alloys may be used. The new concept is based on a helix-shaped tubular heat-exchanger reactor designed by ICI Caldaie (ICI) fitted with CATs proprietary wire-mesh catalyst. The concept has currently been evaluated for production rates between 0.5 and 30 nm3/hr of hydrogen and this report describes a detailed study of a unit for small-scale CHP-applications. Such systems will involve a fuel processor together with a suitable fuel cell, e.g. a solid-oxide fuel cell (SOFC) or a high-temperature PEFC (HT-PEFC). The evaluations performed in this study indicate stable operation over a wide window of capacities with negligible emissions of hydrocarbons. Since it is possible to operate the redesigned unit at a higher temperature (>900 deg C) than the original Ultraformer unit (750- 800 deg C), the conversion degree is much higher for thermo-dynamical reasons. The redesigned unit contains all necessary structures for vaporization, recuperation, effect supply and gas purification in a highly integrated structure. Furthermore, the unit is equipped with an internal insulation and a cooling jacked to reduce the skin temperature of the unit. The reactor has undergone about 100 full thermal cycles without any thermo-mechanical issues or catalyst degradation. Natural gas and different kerosene qualities have so far been evaluated with respect to conversion degree and possible slip of hydrocarbons. The conversion degree at rated load (100%) was above 99%, which enable us to reach superior efficiencies. If the unit were to be used together with a SOFC, the WGS-step could be omitted, reducing the size and weight further from about 2.2 l

  13. A thorium breeder reactor concept for optimal energy extraction from uranium and thorium

    International Nuclear Information System (INIS)

    Jagannnathan, V.; Lawande, S.V.

    1999-01-01

    An attractive thorium breeder reactor concept has been evolved from simple physics based guidelines for induction of thorium in a major way in an otherwise enriched uranium reactor. D 2 O moderator helps to maximise reactivity for a given enrichment. A relatively higher flux level compared to LWRs offers the advantage of higher rate of 233 U production in thoria rods. Thus fresh thoria clusters consider no feed enrichment. In an equilibrium core, a full batch of pure thoria clusters are loaded during each fuel cycle. They undergo irradiation for about one year duration. By this time they accumulate nearly 70% of the asymptotic stable concentration of 233 U, if they face a flux level of the order of 10 14 n/cm 2 /sec. In the next fuel cycle, these thoria rods in ring cluster form are juxtaposed with the fresh enriched fuel rods, also in ring cluster form. Such integrated fuel assemblies are then irradiated for four or five fuel cycles, at the end of which U as well as Th rods attain a reasonably high burnup of about 30-32 MWD/kg. The core characteristics are quite attractive. The core excess reactivity remains low due to large thoria inventory which makes the net burnup reactivity load to be below 1%. The core is capable of being operated in an annual batch mode of operation like a LWR. The control requirement during power operation is negligible. Xenon over-ride requirement is low and can be managed by partial withdrawal of a few thoria clusters. Void reactivity is nearly zero or negative by the optimum design of the fuel cluster. Reactivity changes due to temperatures of fuel, coolant and moderator are also small. (author)

  14. Review of IVR-ERVC and using flooding concept for application to high power reactor

    International Nuclear Information System (INIS)

    Lee, Min ho; Heo, Hyo; Bang, In Cheol

    2016-01-01

    Accident scope will be limited in the RPV. For example, in case of Fukushima, they have difficulties for cleanup the accident and even catching the location of the melt-through corium. Therefore, IVR-ERVC is the right strategy for mitigation of the severe accident. However, in case of high power reactors, there is a Critical Heat Flux (CHF) problem in its application to high power reactor. If CHF occurred, boiling regime changes from effective nucleate boiling to ineffective film boiling, so temperature of the RPV goes up and finally the RPV fails. To solve the CHF problem, here have been a lot of works for IVR-ERVC. In the point of in-vessel heat transfer, Theofanous suggested risk oriented accident analysis methodology which is a combination of probabilistic and deterministic approach. A lot of experiments have been done using simulants of corium in various experimental apparatus. Their simulants were usually water due to simulate large Rayleigh number and natural circulation of corium. IVR-ERVC concept has been researched for a long time. For in-vessel heat transfer, simulants or real corium was used to get a heat flux distribution to the outer wall. And based on those results, ex-vessel cooling has been researched in various geometry to get cooling limit as CHF. Material flooding is suggested as improvement of ERVC in APR 1400 to secure safety margin for CHF. Regardless of Prandtl number of the flooding material, the focusing effect of heat flux was mitigated; the maximum heat flux was reduced less than half of the maximum heat flux in bare condition

  15. Environmentally assisted cracking in light-water reactors: Semi-annual report, January - June 1997. Volume 24

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1998-04-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1997 to June 1997. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Types 304 and 304L SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle is equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in low-DO, simulated pressurized water reactor environments

  16. Some particular aspects of control in nuclear power reactors; Conception de la surete en france et influence des imperatifs de surete sur la conception des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Vathaire, F de; Vernier, Ph; Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [French] La presente communication propose une synthese de l'experience acquise en France en matiere de surete des reacteurs. Les reacteurs de la filiere graphite-gaz faisant l'objet d'une communication particuliere, on examine ici la surete des autres types de reacteurs etudies en France: - reacteurs eau lourde-gaz, - reacteurs a neutrons rapides, - reacteurs de recherche a eau des types piscines et tank. Les imperatifs de surete propres aux differentes filieres sont developpes, en mettant l'accent sur leur influence sur la conception des reacteurs et sur les limitations de puissance qu'ils entrainent. Les etudes de surete correspondantes sont presentees, en insistant plus particulierement sur les travaux originaux developpes dans ces domaines. On indique notamment les moyens d'essais qui ont ete construits pour ces etudes: le reacteur CABRI, boucle en pile pour essais de depressurisation, boucles hors pile, maquettes, etc. (auteurs)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  18. A reactor/receiver-concept for liquid-phase high-temperature processes

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt-Traub, H.; Hahm, T. [Dortmund Univ. (Germany). Dept. of Chemical Engineering

    1997-12-31

    Besides the conversion of solar light to electricity solar energy can be used directly in photo- and thermochemistry. In the temperature range from 1000 to 2000 K there is a high demand for industrial process heat offering a variety of possibilities for solar thermal applications. Especially in the field of liquid-phase high-temperature processes there are hardly no solar thermal applications which exceed the stage of laboratory experiments. It was therefore the aim of two projects financed by the AG Solar of North Rhine-Westphalia, Germany, to develop concepts for commercial scale solar thermal plants and to judge them economically and ecologically. Some general problems have to be overcome to realize a commercial scale solar thermal plant for liquid-phase processes. The concept developed consists of a heliostat field, a tower reflector and an open receiver with a closed reaction chamber. The feasibility of a solar thermal plant for high-temperature liquid-phase processes has been shown in principle. The projected plant consists of a 4400 m{sup 2} heliostat field, a tower plus reflecting mirrors with a total area of 220 m{sup 2} and an open receiver with a closed annular reaction zone. For temperatures below 1700 K the overall efficiency is high enough to yield energetic amortization times of less than 1 year. For a further improvement and a verification of the calculation a closer look at the reactor/receiver and its heat transfer processes is necessary. This is done by using a mixed strategy of experiments and simulation. First experiments were carried out with a semitransparent salt and an opaque metal. The first stage of the experiments will end during the next weeks and their results have to be compared with the simulation. The simulation will then be extended to transparent melts. The second stage of the experiments which include the reaction chamber will start in 1997. An improvement of the reactor might be achieved using nonimaging concentrators to further

  19. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  20. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  1. Hydraulically driven control rod concept for integral reactors: fluid dynamic simulation and preliminary test

    International Nuclear Information System (INIS)

    Ricotti, M.E.; Cammi, A.; Lombardi, C.; Passoni, M.; Rizzo, C.; Carelli, M.; Colombo, E.

    2003-01-01

    The paper deals with the preliminary study of the Hydraulically Driven Control Rod concept, tailored for PWR control rods (spider type) with hydraulic drive mechanism completely immersed in the primary water. A specific solution suitable for advanced versions of the IRIS integral reactor is under investigation. The configuration of the Hydraulic Control Rod device, made up by an external movable piston and an internal fixed cylinder, is described. After a brief description of the whole control system, particular attention is devoted to the Control Rod characterization via Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior, including dynamic equilibrium and stability properties, has been carried out. Finally, preliminary tests were performed in a low pressure, low temperature, reduced length experimental facility. The results are compared with the dynamic control model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performs correctly, allowing stable dynamic equilibrium positions for the Control Rod and stable behavior during withdrawal and insertion steps. (author)

  2. Design concepts and safety concerns of the small and medium size reactors (SMR)

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Lee, Jae Hun; Kim, Hho Jung

    1998-01-01

    The small and medium size reactors (SMR) and interface facilities such as desalination plant are expected to be located near the population area because of restrictions in transporting the plant products such as fresh water to long distance area. To protect the public around the plant facility from the possible release of radioactive materials, the design development of the SMR is focusing on an enhancement of the safety and reliability as well as the economics. In this study, the major safety concepts of the SMR designs significantly different from the current PWR designs are investigated and the safety concerns applicable to the integrated SMR design of Korea (called SMART), were identified. Those safety issues include the use of proven technology, application of strengthening defense in depth, event categorization and selection, simplification of emergency planning, determination of accident source terms and so on. The efforts to resolve the safety concerns in the design stage will provide an improvement of the safety of the SMART design

  3. A Metal Fuel Core Concept for 1000 MWt Advanced Burner Reactor

    International Nuclear Information System (INIS)

    Yang, W.S.; Kim, T.K.; Grandy, C.

    2007-01-01

    This paper describes the core design and performance characteristics of a metal fuel core concept for a 1000 MWt Advanced Burner Reactor. A ternary metal fuel form of U-TRU-Zr was assumed with weapons grade plutonium feed for the startup core and TRU recovered from LWR spent fuel for the recycled equilibrium core. A compact burner core was developed by trade-off between the burnup reactivity loss and TRU conversion ratio, with a fixed cycle length of one-year. In the startup core, the average TRU enrichment is 15.5%, the TRU conversion ratio is 0.81, and the burnup reactivity loss over a cycle is 3.6% Δk. The heavy metal and TRU inventories are 13.1 and 2.0 metric tons, respectively. The average discharge burnup is 93 MWd/kg, and the TRU consumption rate is 55.5 kg/year. For the recycled equilibrium core, the average TRU enrichment is 22.1 %, the TRU conversion ratio is 0.73, and the burnup reactivity loss is 2.2% Δk. The TRU inventory and consumption rate are 2.9 metric tons and 81.6 kg/year, respectively. The evaluated reactivity coefficients provide sufficient negative feedbacks. The control systems provide shutdown margins that are more than adequate. The integral reactivity parameters for quasi-static reactivity balance analysis indicate favorable passive safety features, although detailed safety analyses are required to verify passive safety behavior. (authors)

  4. Design study on evaluation for power conversion system concepts in high temperature gas cooled reactor with gas turbine

    International Nuclear Information System (INIS)

    Minatsuki, Isao; Mizokami, Yorikata

    2007-01-01

    The design studies on High Temperature Gas Cooled Reactor with Gas Turbine (HTGR-GT) have been performed, which were mainly promoted by Japan Atomic Energy Agency (JAEA) and supported by fabricators in Japan. HTGR-GT plant feature is almost determined by selection of power conversion system concepts. Therefore, plant design philosophy is observed characteristically in selection of them. This paper describes the evaluation and analysis of the essential concepts of the HTGR-GT power conversion system through the investigations based on our experiences and engineering knowledge as a fabricator. As a result, the following concepts were evaluated that have advantages against other competitive one, such as the horizontal turbo machine rotor, the turbo machine in an individual vessel, the turbo machine with single shaft, the generator inside the power conversion vessel, and the power conversion system cycle with an intercooler. The results of the study can contribute as reference data when the concepts will be selected. Furthermore, we addressed reasonableness about the concept selection of the Gas Turbine High Temperature Reactor GTHTR300 power conversion system, which has been promoted by JAEA. As a conclusion, we recognized the GTHTR300 would be one of the most promising concepts for commercialization in near future. (author)

  5. A Novel Molten Salt Reactor Concept to Implement the Multi-Step Time-Scheduled Transmutation Strategy

    International Nuclear Information System (INIS)

    Csom, Gyula; Feher, Sandor; Szieberthj, Mate

    2002-01-01

    Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutation capabilities are studied. The goal is the development of a transmutation technique along with a device implementing it, which yield higher transmutation efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named 'multi-region' MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a 'conventional' MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on. (authors)

  6. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

    1988-01-01

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

  7. The Texts of the Instruments Concerning the Agency's Assistance to Indonesia for the Continuation of a Research Reactor Project

    International Nuclear Information System (INIS)

    1970-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Indonesia and the United States of America, and of the Project Agreement between the Agency and the Government of Indonesia concerning the Agency's assistance to that Government for the continuation of a research reactor project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 19 December 1969.

  8. The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-01-24

    The texts of the Supply Agreement between the Agency, the Government of Finland and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Finland, in connection with the Agency's assistance to the Government of Finland in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These agreements entered into force on 30 December 1960.

  9. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1962-03-22

    The texts of the Supply Agreement between the Agency, the Government of Pakistan and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan, in connection with the Agency's assistance to the Government of Pakistan-in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 5 March 1962.

  10. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-11-24

    The texts of the Supply Agreement between the Agency, the Government of Yugoslavia and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Yugoslavia, in connection with the Agency's assistance to the Government of Yugoslavia in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 4 October 1961.

  11. The Texts of the Instruments Concerning the Agency's Assistance to Indonesia for the Continuation of a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1970-02-08

    The texts of the Supply Agreement between the Agency and the Governments of Indonesia and the United States of America, and of the Project Agreement between the Agency and the Government of Indonesia concerning the Agency's assistance to that Government for the continuation of a research reactor project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 19 December 1969.

  12. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-01-30

    The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, in connection with the Agency's assistance to that Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 18 December 1963.

  13. The Texts of the Instruments connected with the Agency's Assistance to Iran in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1967-08-31

    The texts of the Supply Agreement between the Agency and the Governments of Iran and the United States of America, and of the Project Agreement between the Agency and the Government of Iran, connected with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. The Agreements entered into force on 7 June and 10 May 1967 respectively.

  14. The Texts of the Instruments Concerning the Agency's Assistance to Chile for the Establishment of a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1970-02-09

    The texts of the Supply Agreement between the Agency and the Governments of Chile and the United States of America, and of the Project Agreement between the Agency and the Government of Chile concerning the Agency's assistance to that Government for the establishment of a research reactor project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 19 December 1969.

  15. The Texts of the Instruments Concerning the Agency's Assistance to Venezuela for the Continuation of a Research Reactor Project

    International Nuclear Information System (INIS)

    1976-01-01

    The texts of the Title Transfer Agreement between the Agency and the Governments of the United States of America and Venezuela, and of the Project Agreement between the Agency and the Government of Venezuela concerning the Agency's assistance to that Government for the continuation of a research reactor project, are reproduced herein for the information of all Members. Both Agreements entered into force on 7 November 1975, pursuant to Articles IV and X respectively.

  16. The Texts of the Instruments Concerning the Agency's Assistance to Chile for the Establishment of a Research Reactor Project

    International Nuclear Information System (INIS)

    1970-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Chile and the United States of America, and of the Project Agreement between the Agency and the Government of Chile concerning the Agency's assistance to that Government for the establishment of a research reactor project, are reproduced in this document for the information of all Members. Both Agreements entered into force on 19 December 1969.

  17. The Texts of the Instruments connected with the Agency's assistance to Uruguay in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1965-01-01

    The texts of the Supply Agreement between the Agency and the Governments of the United States of America and Uruguay, and of the Project Agreement between the Agency and the Government of Uruguay, in connection with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 24 September 1965

  18. The texts of the instruments concerning the Agency's assistance to Romania for the establishment of a research reactor project

    International Nuclear Information System (INIS)

    1991-11-01

    As a sequel to the assistance which the Agency has provided to the Government of Romania in connection with a research reactor project (IAEA-INFCIRC/206, 206 Adds.1 and 2 and Mod.2), the Agency and the Governments of Romania and the United States of America, on 14 June 1991, concluded a Fourth Supply Agreement. The text of that Agreement, which was approved by the Board of Governors on 14 June 1991, is reproduced in this document

  19. The texts of the instruments concerning the Agency's assistance to Romania for the establishment of a research reactor project

    International Nuclear Information System (INIS)

    1991-08-01

    On June 15, 1990, the Agency and the Government of Romania signed a letter constituting an agreement amending the Project Agreement (IAEA-INFCIRC/206, Part II) concluded in connection with the Agency's assistance to Romania for the establishment of a research reactor project. The text of that letter, which was approved by the Board of Governors on June 15, 1990, is reproduced in this document. 1 tab

  20. The Texts of the Instruments connected with the Agency's assistance to Uruguay in Establishing a Research Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-09

    The texts of the Supply Agreement between the Agency and the Governments of the United States of America and Uruguay, and of the Project Agreement between the Agency and the Government of Uruguay, in connection with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 24 September 1965.

  1. The Texts of the Instruments concerning the Agency's Assistance to Greece for the continuation of a Research Reactor Project

    International Nuclear Information System (INIS)

    1972-01-01

    The texts of the Title Transfer Agreement between the Agency and the Governments of Greece and the United States of America, and of the Project Agreement between the Agency and the Government of Greece concerning the Agency's assistance to that Government for the continuation of a research reactor project, are reproduced herein for the information of all Members. Both Agreements entered into force on 1 March 1972

  2. The Texts of the Instruments connected with the Agency's Assistance to Pakistan in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1962-01-01

    The texts of the Supply Agreement between the Agency, the Government of Pakistan and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Pakistan, in connection with the Agency's assistance to the Government of Pakistan-in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 5 March 1962

  3. The Texts of the Instruments connected with the Agency's Assistance to Finland in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1961-01-01

    The texts of the Supply Agreement between the Agency, the Government of Finland and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Finland, in connection with the Agency's assistance to the Government of Finland in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These agreements entered into force on 30 December 1960

  4. The Texts of the Instruments connected with the Agency's Assistance to Iran in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1967-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Iran and the United States of America, and of the Project Agreement between the Agency and the Government of Iran, connected with the Agency's assistance to the latter Government in establishing a research reactor project, are reproduced in this document for the information of all Members. The Agreements entered into force on 7 June and 10 May 1967 respectively

  5. The Texts of the Instruments connected with the Agency's Assistance to Yugoslavia in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1961-01-01

    The texts of the Supply Agreement between the Agency, the Government of Yugoslavia and the Government of the United States of America, and of the Project Agreement between the Agency and the Government of Yugoslavia, in connection with the Agency's assistance to the Government of Yugoslavia in establishing a research reactor project, are reproduced in this document for the information of all Members of the Agency. These Agreements entered into force on 4 October 1961

  6. The Texts of the Instruments connected with the Agency's Assistance to Mexico in Establishing a Research Reactor Project

    International Nuclear Information System (INIS)

    1964-01-01

    The texts of the Supply Agreement between the Agency and the Governments of Mexico and the United States of America, and of the Project Agreement between the Agency and the Government of Mexico, in connection with the Agency's assistance to that Government in establishing a research reactor project, are reproduced in this document for the information of all Members. These Agreements entered into force on 18 December 1963

  7. The Text of the Instrument concerning the Agency's Assistance to Mexico for the Establishment of a Training Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-05

    The texts of the Agreement between the Agency the the Governments of the Federal Republic of Germany and Mexico concerning the Agency's assistance for the establishment of a training reactor project in Mexico, and of a letter relating thereto which the Resident Representative of the Federal Republic of Germany addressed to the Director General, are reproduced herein for the information of all Members. The Agreement entered into force on 21 December 1971.

  8. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    International Nuclear Information System (INIS)

    Murata, Takashi; Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-01-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  9. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-12-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  10. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    Science.gov (United States)

    Sibille, Laurent; Dominguez, Jesus A.

    2012-01-01

    The technology of direct electrolysis of molten lunar regolith to produce oxygen and molten metal alloys has progressed greatly in the last few years. The development of long-lasting inert anodes and cathode designs as well as techniques for the removal of molten products from the reactor has been demonstrated. The containment of chemically aggressive oxide and metal melts is very difficult at the operating temperatures ca. 1600 C. Containing the molten oxides in a regolith shell can solve this technical issue and can be achieved by designing a Joule-heated (sometimes called 'self-heating') reactor in which the electrolytic currents generate enough Joule heat to create a molten bath. Solutions obtained by multiphysics modeling allow the identification of the critical dimensions of concept reactors.

  11. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C; Kremer, H [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P; Hupa, M [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1998-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  12. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

  13. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  14. Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

    International Nuclear Information System (INIS)

    Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

    2012-01-01

    Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

  15. Establishment of design concept of large capacity passive reactor KP1000 and performance evaluation of safety system for LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O.; Hwang, Young Dong; Kim, Young In; Chang, Moon Hee

    1997-03-01

    This study was performed to establish the design concepts and to evaluate the performance of safety features of large capacity passive reactor (1000 MWe grade). The design concepts of the large capacity passive reactor `KP1000` were established to generate 1000 MW electric power based on the AP600 of Westinghouse by increasing the number of reactor coolant loop and by increasing the size of reactor internals/core. To implement the analysis of the LBLOCA for KP1000, various kinds of computer codes being considered, it was concluded that RELAP5 was the most appropriate one in availability and operations in present situation. By the analysis of the computer code `RELAP5/Mod3.2.1.2`, following conclusions were derived as described below. First, by spectrum analysis of the discharge factor of the berak part, the most conservative discharge factor C{sub D}=1.2 and the PCT value of KP1000 was 1254F, which is slightly higher than the value of AP600 but is much less than the existing active reactor `Kori 3 and 4` where blowdown PCT value is 1693.4 deg F and reflooding PCT is 1918.4 deg F. Second, after the 200 seconds from the initiation of LBLOCA, IRWST water was supplied in a stable state and the maximum temperature of clad were maintained in a saturated condition. Therefore, it was concluded that the passive safety features of KP1000 keep reactor core from being damaged for large break LOCA. (author). 11 refs., 28 tabs., 37 figs.

  16. Design of subjects training on reactor simulator and feasibility study - toward the empirical evaluation of interface design concept

    International Nuclear Information System (INIS)

    Yamaguchi, Y.; Furukawa, H.; Tanabe, F.

    1998-01-01

    On-going JAERI's project for empirical evaluation of the ecological interface design concept was first described. The empirical evaluation is planned to be proceeded through three consecutive steps; designing and actual implementation of the interface on reactor simulator, verification of the interface created, and the validation by the simulator experiment. For conducting the project, three different experimental resources are prerequisite, that are, data analysis method for identifying the operator's strategies, experimental facility including reactor simulator, and experimental subjects or subjects training method. Among the three experimental resources, subjects training method was recently designed and a simulator experiment was earned out in order to examine the feasibility of the designed training method. From the experiment and analysis of the experimental records, we could conclude that it is feasible that the experimental subjects having an appropriate technical basis can gain the sufficient competence for evaluation work of the interface design concept by using the training method designed. (author)

  17. Improved best estimate plus uncertainty methodology, including advanced validation concepts, to license evolving nuclear reactors

    International Nuclear Information System (INIS)

    Unal, C.; Williams, B.; Hemez, F.; Atamturktur, S.H.; McClure, P.

    2011-01-01

    Research highlights: → The best estimate plus uncertainty methodology (BEPU) is one option in the licensing of nuclear reactors. → The challenges for extending the BEPU method for fuel qualification for an advanced reactor fuel are primarily driven by schedule, the need for data, and the sufficiency of the data. → In this paper we develop an extended BEPU methodology that can potentially be used to address these new challenges in the design and licensing of advanced nuclear reactors. → The main components of the proposed methodology are verification, validation, calibration, and uncertainty quantification. → The methodology includes a formalism to quantify an adequate level of validation (predictive maturity) with respect to existing data, so that required new testing can be minimized, saving cost by demonstrating that further testing will not enhance the quality of the predictive tools. - Abstract: Many evolving nuclear energy technologies use advanced predictive multiscale, multiphysics modeling and simulation (M and S) capabilities to reduce the cost and schedule of design and licensing. Historically, the role of experiments has been as a primary tool for the design and understanding of nuclear system behavior, while M and S played the subordinate role of supporting experiments. In the new era of multiscale, multiphysics computational-based technology development, this role has been reversed. The experiments will still be needed, but they will be performed at different scales to calibrate and validate the models leading to predictive simulations for design and licensing. Minimizing the required number of validation experiments produces cost and time savings. The use of multiscale, multiphysics models introduces challenges in validating these predictive tools - traditional methodologies will have to be modified to address these challenges. This paper gives the basic aspects of a methodology that can potentially be used to address these new challenges in

  18. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  19. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    International Nuclear Information System (INIS)

    Zdarek, J.; Pecinka, L.

    1997-01-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors

  20. Clinical and molecular analyses of Beckwith-Wiedemann syndrome: Comparison between spontaneous conception and assisted reproduction techniques.

    Science.gov (United States)

    Tenorio, Jair; Romanelli, Valeria; Martin-Trujillo, Alex; Fernández, García-Moya; Segovia, Mabel; Perandones, Claudia; Pérez Jurado, Luis A; Esteller, Manel; Fraga, Mario; Arias, Pedro; Gordo, Gema; Dapía, Irene; Mena, Rocío; Palomares, María; Pérez de Nanclares, Guiomar; Nevado, Julián; García-Miñaur, Sixto; Santos-Simarro, Fernando; Martinez-Glez, Víctor; Vallespín, Elena; Monk, David; Lapunzina, Pablo

    2016-10-01

    Beckwith-Wiedemann syndrome (BWS) is an overgrowth syndrome characterized by an excessive prenatal and postnatal growth, macrosomia, macroglossia, and hemihyperplasia. The molecular basis of this syndrome is complex and heterogeneous, involving genes located at 11p15.5. BWS is correlated with assisted reproductive techniques. BWS in individuals born following assisted reproductive techniques has been found to occur four to nine times higher compared to children with to BWS born after spontaneous conception. Here, we report a series of 187 patients with to BWS born either after assisted reproductive techniques or conceived naturally. Eighty-eight percent of BWS patients born via assisted reproductive techniques had hypomethylation of KCNQ1OT1:TSS-DMR in comparison with 49% for patients with BWS conceived naturally. None of the patients with BWS born via assisted reproductive techniques had hypermethylation of H19/IGF2:IG-DMR, neither CDKN1 C mutations nor patUPD11. We did not find differences in the frequency of multi-locus imprinting disturbances between groups. Patients with BWS born via assisted reproductive techniques had an increased frequency of advanced bone age, congenital heart disease, and decreased frequency of earlobe anomalies but these differences may be explained by the different molecular background compared to those with BWS and spontaneous fertilization. We conclude there is a correlation of the molecular etiology of BWS with the type of conception. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  1. Safety studies dedicated to molten salt reactors with a fast neutron spectrum and operated in the Thorium fuel cycle - Innovative concept of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Brovchenko, Mariya

    2013-01-01

    The nuclear reactors of the 4. generation must allow an optimized use of natural resources, while performing at a high safety level. The framework of this thesis is the deployment study of one of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellent safety is an ultimate requirement of the nuclear energy deployment, so it is important to raise this question at the current early stage of the MSFR concept development. This concept was the subject of a neutronic tool benchmark within a European project EVOL. Definition, calculations and results analyses were performed during this thesis. Comparisons of static neutronic and burn-up calculations, performed by the project participants, concluded to a good agreement between the different codes and methods used and pointed out the sensibility of the nuclear database choice on the results. With the aim of safety analysis of the MSFR, the decay heat was studied in detail. The tool used for the decay heat calculation was developed and validated, to finally evaluate the decay heat in the reactor. The decay heat source presented in different zones was quantified, concluding to a high importance of the cooling of the fuel salt and the bubbling system enclosing a part of the fission products. The safety analysis methodology was also studied in this thesis. Even if the safety principles are directly transposable to the MSFR, the precise recommendations are not. This is due to the specificity of the design that relies on the liquid state of the fuel, on the reprocessing systems located in the reactor and the embryonic stage of the design. First, a preliminary transposition work of some criteria to the MSFR design was realized, resulting amongst other things in a list of accidental scenarios particular for MSFR. Finally, a preliminary physical study of some types of accidental scenarios was performed, that can be used as a basis for further analyses with more sophisticated tools. (author) [fr

  2. Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

    International Nuclear Information System (INIS)

    McDonald, M.; Colton, A.; Pencer, J.

    2015-01-01

    The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

  3. Conflicting modes of reasoning in the Assisted Migration debate: a concept mapping analysis

    Science.gov (United States)

    Klenk, N. L.; Larson, B.

    2011-12-01

    A growing body of literature in biodiversity conservation and forestry has developed around the risks and benefits of the Assisted Migration (AM) of species, including recommendations for planning and management. However, despite years of academic debate, general consensus has yet to be reached between the proponents and the opponents of AM as a policy option. We hypothesize that the continued impasse arises out of fundamentally conflicting value judgments. Using a concept mapping technique, we analyzed reasons for and against AM, including the use of scientific evidence and the nature of the values and ethical norms that shape the modes of reasoning in the debate. Our results indicate the presence of a diversity of ethical arguments in addition to the standard precautionary argument and pragmatic reasoning. We further discovered that different kinds of scientific arguments are used by proponents versus opponents of AM: the former rely mostly on detailed biological and ecological facts about species most-at-risk under climate change, while the latter focus on broader ecological theories. Our analysis suggests little dispute over the scientific foundations of the debate. Instead, we suggest that the main barrier to consensus is the advocacy of fundamental values, which are a matter of personal choice, and thus not likely to be changed. One way out of this impasse is a pragmatic mode of reasoning, which eschews the debate on fundamental values and evaluates the means and ends of AM in a case-by-case approach.

  4. Effectiveness of Using Computer-Assisted Supplementary Instruction for Teaching the Mole Concept

    Science.gov (United States)

    Yalçinalp, Serpil; Geban, Ömer; Özkan, Ilker

    This study examined the effect of computer-assisted instruction (CAI), used as a problem-solving supplement to classroom instruction, on students' understanding of chemical formulas and mole concept, their attitudes toward chemistry subjects, and CAI. The objective was to assess the effectiveness of CAI over recitation hours when both teaching methods were used as a supplement to the traditional chemistry instruction. We randomly selected two classes in a secondary school. Each teaching strategy was randomly assigned to one class. The experimental group received supplementary instruction delivered via CAI, while the control group received similar instruction through recitation hours. The data were analyzed using two-way analysis of variance and t-test. It was found that the students who used the CAI accompanied with lectures scored significantly higher than those who attended recitation hours, in terms of school subject achievement in chemistry and attitudes toward chemistry subjects. In addition, there was a significant improvement in the attitudes of students in the experimental group toward the use of computers in a chemistry course. There was no significant difference between the performances of females versus males in each treatment group.Received: 26 April 1994; Revised: 6 April 1995;

  5. Emotion-Aware Assistive System for Humanistic Care Based on the Orange Computing Concept

    Directory of Open Access Journals (Sweden)

    Jhing-Fa Wang

    2012-01-01

    Full Text Available Mental care has become crucial with the rapid growth of economy and technology. However, recent movements, such as green technologies, place more emphasis on environmental issues than on mental care. Therefore, this study presents an emerging technology called orange computing for mental care applications. Orange computing refers to health, happiness, and physiopsychological care computing, which focuses on designing algorithms and systems for enhancing body and mind balance. The representative color of orange computing originates from a harmonic fusion of passion, love, happiness, and warmth. A case study on a human-machine interactive and assistive system for emotion care was conducted in this study to demonstrate the concept of orange computing. The system can detect emotional states of users by analyzing their facial expressions, emotional speech, and laughter in a ubiquitous environment. In addition, the system can provide corresponding feedback to users according to the results. Experimental results show that the system can achieve an accurate audiovisual recognition rate of 81.8% on average, thereby demonstrating the feasibility of the system. Compared with traditional questionnaire-based approaches, the proposed system can offer real-time analysis of emotional status more efficiently.

  6. Computer-assisted generation of individual training concepts for advanced education in manufacturing metrology

    International Nuclear Information System (INIS)

    Werner, Teresa; Weckenmann, Albert

    2010-01-01

    Due to increasing requirements on the accuracy and reproducibility of measurement results together with a rapid development of novel technologies for the execution of measurements, there is a high demand for adequately qualified metrologists. Accordingly, a variety of training offers are provided by machine manufacturers, universities and other institutions. Yet, for an interested learner it is very difficult to define an optimal training schedule for his/her individual demands. Therefore, a computer-based assistance tool is developed to support a demand-responsive scheduling of training. Based on the difference between the actual and intended competence profile and under consideration of amending requirements, an optimally customized qualification concept is derived. For this, available training offers are categorized according to different dimensions: regarding contents of the course, but also intended target groups, focus of the imparted competences, implemented methods of learning and teaching, expected constraints for learning and necessary preknowledge. After completing a course, the achieved competences and the transferability of gathered knowledge are evaluated. Based on the results, recommendations for amending measures of learning are provided. Thus, a customized qualification for manufacturing metrology is facilitated, adapted to the specific needs and constraints of each individual learner

  7. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  8. The dynomak: An advanced spheromak reactor concept with imposed-dynamo current drive and next-generation nuclear power technologies

    Energy Technology Data Exchange (ETDEWEB)

    Sutherland, D.A., E-mail: das1990@uw.edu; Jarboe, T.R.; Morgan, K.D.; Pfaff, M.; Lavine, E.S.; Kamikawa, Y.; Hughes, M.; Andrist, P.; Marklin, G.; Nelson, B.A.

    2014-04-15

    A high-β spheromak reactor concept has been formulated with an estimated overnight capital cost that is competitive with conventional power sources. This reactor concept utilizes recently discovered imposed-dynamo current drive (IDCD) and a molten salt (FLiBe) blanket system for first wall cooling, neutron moderation and tritium breeding. Currently available materials and ITER-developed cryogenic pumping systems were implemented in this concept from the basis of technological feasibility. A tritium breeding ratio (TBR) of greater than 1.1 has been calculated using a Monte Carlo N-Particle (MCNP5) neutron transport simulation. High temperature superconducting tapes (YBCO) were used for the equilibrium coil set, substantially reducing the recirculating power fraction when compared to previous spheromak reactor studies. Using zirconium hydride for neutron shielding, a limiting equilibrium coil lifetime of at least thirty full-power years has been achieved. The primary FLiBe loop was coupled to a supercritical carbon dioxide Brayton cycle due to attractive economics and high thermal efficiencies. With these advancements, an electrical output of 1000 MW from a thermal output of 2486 MW was achieved, yielding an overall plant efficiency of approximately 40%.

  9. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    International Nuclear Information System (INIS)

    Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jaek Wan

    2012-01-01

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security

  10. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Jin Soo; Heo, Gyun Young [Kyung Hee University, Seoul (Korea, Republic of); Son, Han Seong [Joongbu Univ., Chungnam (Korea, Republic of); Kim, Young Ki; Park, Jaek Wan [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security.

  11. 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.)

  12. 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.)

  13. Proposals of new basic concepts on safety and radioactive waste and of new High Temperature Gas-cooled Reactor based on these basic concepts

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    2016-01-01

    Highlights: • The author proposed new basic concepts on safety and radioactive waste. • A principle of ‘continue confining’ to realize the basic concept on safety is also proposed. • It is indicated that only a HTGR can attain the conditions required from the principle. • Technologies to realize the basic concept on radioactive waste are also discussed. • A New HTGR system based on the new basic concepts is proposed. - Abstract: A new basic concept on safety of ‘Not causing any serious catastrophe by any means’ and a new basic concept on radioactive waste of ‘Not returning any waste that possibly affects the environment’ are proposed in the present study, aiming at nuclear power plants which everybody can accept, in consideration of the serious catastrophe that happened at Fukushima Japan in 2011. These new basic concepts can be found to be valid in comparison with basic concepts on safety and waste in other industries. The principle to realize the new basic concept on safety is, as known well as the inherent safety, to use physical phenomena such as Doppler Effect and so on which never fail to work even if all equipment and facilities for safety lose their functions. In the present study, physical phenomena are used to ‘continue confining’, rather than ‘confine’, because the consequence of emission of radioactive substances to the environment cannot be mitigated. To ‘continue confining’ is meant to apply natural correction to fulfill inherent safety function. Fission products must be detoxified to realize the new basic concept on radioactive waste, aiming at the final processing and disposal of radioactive wastes as same as that in the other wastes such as PCB, together with much efforts not to produce radioactive wastes and to reduce their volume nevertheless if they are emitted. Technology development on the detoxification is one of the most important subjects. A new High Temperature Gas-cooled Reactor, namely the New HTGR

  14. Proposals of new basic concepts on safety and radioactive waste and of new High Temperature Gas-cooled Reactor based on these basic concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro, E-mail: ogawa.masuro@jaea.go.jp

    2016-11-15

    Highlights: • The author proposed new basic concepts on safety and radioactive waste. • A principle of ‘continue confining’ to realize the basic concept on safety is also proposed. • It is indicated that only a HTGR can attain the conditions required from the principle. • Technologies to realize the basic concept on radioactive waste are also discussed. • A New HTGR system based on the new basic concepts is proposed. - Abstract: A new basic concept on safety of ‘Not causing any serious catastrophe by any means’ and a new basic concept on radioactive waste of ‘Not returning any waste that possibly affects the environment’ are proposed in the present study, aiming at nuclear power plants which everybody can accept, in consideration of the serious catastrophe that happened at Fukushima Japan in 2011. These new basic concepts can be found to be valid in comparison with basic concepts on safety and waste in other industries. The principle to realize the new basic concept on safety is, as known well as the inherent safety, to use physical phenomena such as Doppler Effect and so on which never fail to work even if all equipment and facilities for safety lose their functions. In the present study, physical phenomena are used to ‘continue confining’, rather than ‘confine’, because the consequence of emission of radioactive substances to the environment cannot be mitigated. To ‘continue confining’ is meant to apply natural correction to fulfill inherent safety function. Fission products must be detoxified to realize the new basic concept on radioactive waste, aiming at the final processing and disposal of radioactive wastes as same as that in the other wastes such as PCB, together with much efforts not to produce radioactive wastes and to reduce their volume nevertheless if they are emitted. Technology development on the detoxification is one of the most important subjects. A new High Temperature Gas-cooled Reactor, namely the New HTGR

  15. Coming To Know: The Role of the Concept Map--Mirror, Assistant, Master?

    Science.gov (United States)

    McAleese, Ray

    This paper explains the process of creating and managing concept maps, using reflection as a focus for its argument. Section 1, What is a Concept Map?, highlights the background and definition of concept mapping, explains how maps signify virtual conceptual structures, looks at structural knowledge, provides an example of a concept map, and…

  16. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  17. Evaluation for reasonableness of power conversion system concepts in the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Minatsuki, I.; Mizokami, Y.

    2007-01-01

    The conceptual design study for the Gas Turbine High Temperature Reactor (GTHTR300) was completed in 2004. In GTHTR300, SECO (Simple, Economical Competitiveness and Originality) is advocated as design philosophy in order to minimize technical and economical requirement. Furthermore the design of the GTHTR300 was developed with reflecting various view points from utilities, manufacturers and research organizations. In GTHTR300, the horizontal turbo machine rotor, the turbo machine in a separated vessel, the turbo machine with single rotor, the generator inside the power conversion vessel, and the power conversion system without inter-coolers were selected as major power conversion system concepts. This paper describes the investigation and analysis about the major concepts of GTHTR300 power conversion system in order to evaluate reasonableness of GTHTR300 design approach and acceptability with using experience and engineering knowledge of Mitsubishi Heavy Industries, Ltd., which were accumulated through the activities of HTGR-GT and HTTR (High Temperature Engineering Test Reactor) designing, manufacturing, fabricating and testing. From the result of the evaluation, it was concluded that the selection of each concept in GTHTR300 was reasonable as based on the original design philosophy SECO. As a conclusion, we expect the GTHTR300 to become one of the most promising concepts for commercialization in near future. (authors)

  18. Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts

    International Nuclear Information System (INIS)

    Miller, Laurence F.; Khuram Khan, M.; Williams, Wesley; Mynatt, F.R.

    2002-01-01

    This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model

  19. Safety enhancement concept for NPP of new generation with VVER reactors

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Kukhtevich, I.; Semashko, S.; Svetlov, S.; Solodovnikov, A.

    2004-01-01

    With the present day conditions, in order to successfully promote new NPP designs in the electric power markets, it is necessary to ensure enhanced technical/economic performances provided that international safety requirements are properly adhered to. When compared with high-powered nuclear power plants, NPP VVER-640 design (medium powered) possesses a number of advantages for the regions with undeveloped energy systems. Reduced specific energy intensity of the core adopted in this type of reactor allows to ensure the emergency cooldown of the reactor plant by passive means and to minimize the 'human factor' risk and external effects and provide sound substantiations as to how to retain corium inside RPV in case of severe accidents. At the same time, high-powered NPPs seem to be promising for regions with developed energy systems. Among such designs, NPP VVER-1000 and VVER-1500 designs are the most desirable. Configuration of new generation NPP with VVER-1500 is to be selected based on the gained experience in designing NPPs of previous generations considering the latest safety requirements and situation in the domestic and global energy markets for the time being and in the short run. Recent IAEA publications and latest EUR requirements insist that the following key safety indices should be established for new NPP designs: - aggregated frequency of core melting is 10 -6 (1/year); - frequency of maximum accident release is 10 -7 (1/year). To meet the aforementioned criteria, it is necessary to implement some safety assurance principles recommended by IAEA (in-depth defence, single failure, redundancy, diversity, etc.), application of deterministic and probabilistic methods for selection of safety assurance activities and means and use of reasonable combination of active and passive systems. Application of VVER-640 concept to high-powered NPPs seems to be a formidable task due to a number of reasons, namely, it is quite difficult to carry out cooldown process

  20. The Concept of the Use of the Marine Reactor Plant in Small Electric Grids

    International Nuclear Information System (INIS)

    Khlopkin, N.; Makarov, V.; Pologikh, B.

    2002-01-01

    In report some aspects of the using marine nuclear reactor are considered for provision of need small non-interconnected power systems, as well as separate settlements and the mining enterprises disposed in regions with a undeveloped infrastructure. Recently for these purposes it is offered to use the nuclear small modular power plants. The required plant power for small electric grids lies within from 1 to several tens of MWe. Module can be collected and tested on machine-building plant, and then delivered in ready type to the working place on some transport, for instance, a barge. Through determined time it's possible to transport a module to the repair shop and also to the point of storage after the end of operation. Marine nuclear reactors on their powers, compactness, mass and size are ideal prototypes for creation of such modules. For instance, building at present floating power unit, intended for functioning in region of the Russian North, based on using reactor plants of nuclear icebreakers. Reliability and safety of the ship reactor are confirmed by their trouble-free operation during approximately 180 reactors-years. Unlike big stationary nuclear plant, working in base mode, power unit with marine reactor wholly capable to work in mode of the loading following. In contrast with reactor of nuclear icebreaker, advisable to increase the core lifetime and to reduce the enrichment of the uranium. This requires more uranium capacity fuel compositions and design of the core. In particular, possible transition from traditional for ship reactor of the channel core to cassette design. Other directions of evolution of the ship reactors, not touching the basic constructive decisions verified by practice, but promoting development of properties of self-security of plant are possible. Among such directions is reduction volumetric power density of a core. (author)

  1. Can the acceptance of nuclear reactors be raised by a simpler, more transparent safety concept employing improved containments?

    International Nuclear Information System (INIS)

    Krieg, R.

    1993-01-01

    Sociological and psychological findings are presented which describe problems society faces with risky technologies. It turns out that for a small but influential group of the society not only the risk, but also the transparency of the technology and the engineered safeguards are important. For this group an improvement of the transparency should raise the acceptance of the technology. However, for the majority of the public, broad discussions about a different more transparent safety concept could generate a feeling of insecurity, especially concerning existing nuclear power plants. In order to improve the transparency in engineered safeguards of nuclear reactors, four basic principles are introduced. Nuclear containments, which remain intact after failure of all emergency cooling systems resulting in a core meltdown accident with steam and hydrogen explosions, are in line with these principles. Such containments are under investigation in Germany right now. It is pointed out that the public should be informed prudently about this work to improve the containments, and the required transparency of the engineered safeguards should be underlined as a major goal. Vague descriptions, providing only superficially information about safer reactors, could easily cause misunderstandings. If one goes along this line a simpler and more transparent safety concept employing improved containments will have the potential to raise the acceptance of nuclear reactors. (orig.)

  2. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

    Science.gov (United States)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1998-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

  3. Principle of an operational complexity index for the characterization of the human factor relevance of future reactors concepts

    International Nuclear Information System (INIS)

    Papin, Bernard

    2004-01-01

    With the increasing reliability of the modern technological systems, the human contribution to the global risk in the operation of industrial systems is becoming more and more significant : in the nuclear reactor operation for example, a recent PSA estimation of this contribution is about 25% of the risk of core melting, all situations considered. This urges the designers of future nuclear reactors to consider the minimisation of this Human Factor (HF) contribution, at the very early stage of their design : the experience feedback shows that this is indeed at this stage that the fundamental design options, impacting the most the human reliability in operation, are fixed. The problem is that at these early design stages, it is also quite impossible to apply formal human reliability methods to support this HF optimisation, while the precise operating conditions of the reactor are not yet known in enough details. In this paper, another approach of the HF evaluation during the design, based on the functional and operational complexity assessment, is proposed. As an illustration, this approach is used to compare various concepts of Pressurized Water Reactors from the point of view of the Human Factor relevance. (Author)

  4. CO_2 capture with solid sorbent: CFD model of an innovative reactor concept

    International Nuclear Information System (INIS)

    Barelli, L.; Bidini, G.; Gallorini, F.

    2016-01-01

    Highlights: • A new reactor solution based on rotating fixed beds was presented. • The preliminary design of the reactor was approached. • A CFD model of the reactor, including CO_2 capture kinetic, was developed. • The CFD model is validated with experimental results. • Sorbent exploitation increasing is possible thanks to the new reactor. - Abstract: In future decarbonization scenarios, CCS with particular reference to post-combustion technologies will be an important option also for energy intensive industries. Nevertheless, today CCS systems are rarely installed due to high energy and cost penalties of current technology based on chemical scrubbing with amine solvent. Therefore, innovative solutions based on new/optimized solvents, sorbents, membranes and new process designs, are R&D priorities. Regarding the CO_2 capture through solid sorbents, a new reactor solution based on rotating fixed beds is presented in this paper. In order to design the innovative system, a suitable CFD model was developed considering also the kinetic capture process. The model was validated with experimental results obtained by the authors in previous research activities, showing a potential reduction of energy penalties respect to current technologies. In the future, the model will be used to identify the control logic of the innovative reactor in order to verify improvements in terms of sorbent exploitation and reduction of system energy consumption.

  5. The concept of laparoscopy-assisted pancreatobiliary EUS (LAP-EUS).

    Science.gov (United States)

    Ramesh, Jayapal; Christein, John; Varadarajulu, Shyam

    2013-05-01

    While laparoscopy-assisted endoscopic retrograde cholangiopancreatography can be performed for the diagnosis and management of pancreaticobiliary diseases in patients with Roux-en-Y (RY) anatomy; the technical feasibility of performing laparoscopy-assisted endoscopic ultrasound (EUS) is unknown. In this report, we describe the technique for performing laparoscopy-assisted EUS in two patients with RY gastric bypass anatomy who presented with obstructive jaundice, abnormal liver function tests, and dilated biliary tree. While the examination was normal in one patient, EUS-guided fine needle aspiration of pancreatic head mass revealed adenocarcinoma in the other. Laparoscopic assisted EUS examination including FNA is feasible in Roux-en-Y surgical anatomy.

  6. The text of the instrument concerning the Agency's assistance to Argentina for the establishment of a training reactor project

    International Nuclear Information System (INIS)

    1998-01-01

    The document reproduces the text of the Protocol of 30 August 1996 suspending the application of safeguards under the Agreement between the Agency and the Government of Argentina and the Federal Republic of Germany concerning the Agency's assistance for the establishment of a training reactor project in Argentina, which entered into force on 13 March 1970, in light of the provisions for the application of safeguards pursuant to the quadripartite safeguards agreement between Argentina, Brazil, the Brazilian-Argentine Agency for the Accounting and Control of Nuclear Materials and the IAEA

  7. The texts of the instruments concerning the Agency's assistance to Romania for the establishment of a research reactor project

    International Nuclear Information System (INIS)

    1991-08-01

    As a sequel to the assistance which the Agency has provided to the Government of Romania in connection with a research reactor project (IAEA-INFCIRC/206, IAEA-INFCIRC/206/ Add.1 and IAEA-INFCIRC/206/Mod.2), the Agency, the Government of Romania and the United States of America, on June 15, 1990, signed a letter constituting the Third Supply Agreement. The text of that letter, which was approved by the Board of Governors on June 15, 1990, is reproduced in this document

  8. Interim report on core physics and fuel cycle analysis of the pebble bed reactor power plant concept

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1977-12-01

    Calculations were made to predict the performance of a pebble bed reactor operated in a mode to produce fissile fuel (high conversion or breeding). Both a one pebble design and a design involving large primary feed pebbles and small fertile pebbles were considered. A relatively short residence time of the primary pebbles loaded with 233 U fuel was found to be necessary to achieve a high breeding ratio, but this leads to relatively high fuel costs. A high fissile inventory is associated with a low C/Th ratio and a high thorium loading, causing the doubling time to be long, even though the breeding ratio is high, and the fuel cost of electrical product to be high. Production of 233 U fuel from 235 U feed was studied and performances of the converter and breeder reactor concepts were examined varying the key parameters

  9. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    Science.gov (United States)

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis.

  10. Assisted Reproduction versus Spontaneous Conception: A Comparison of the Developmental Outcomes in Twins

    Science.gov (United States)

    Kelly-Vance, Lisa; Anthis, Kristine S.; Needelman, Howard

    2004-01-01

    The use of assisted reproductive technology is increasing rapidly. Research, although sparse, has resulted in inconsistent findings as to the developmental prognosis for infants conceived by assisted reproductive techniques such as in vitro fertilization and the use of fertility drugs. In the present study, the authors compared twins who were…

  11. Trends in 'poor responder' research: lessons learned from RCTs in assisted conception.

    Science.gov (United States)

    Papathanasiou, Athanasios; Searle, Belinda J; King, Nicole M A; Bhattacharya, Siladitya

    2016-04-01

    A substantial minority of women undergoing IVF will under-respond to controlled ovarian hyperstimulation. These women-so-called 'poor responders'-suffer persistently reduced success rates after IVF. Currently, no single intervention is unanimously accepted as beneficial in overcoming poor ovarian response (POR). This has been supported by the available research on POR, which consists mainly of randomized controlled trials (RCTs ) with an inherent high-risk of bias. The aim of this review was to critically appraise the available experimental trials on POR and provide guidance towards more useful-less wasteful-future research. A comprehensive review was undertaken of RCTs on 'poor responders' published in the last 15 years. Data on various methodological traits as well as important clinical characteristics were extracted from the included studies and summarized, with a view to identifying deficiencies from which lessons can be learned. Based on this analysis, recommendations were provided for further research in this field of assisted conception. We selected and analysed 75 RCTs. A valid, 'low-risk' randomization method was reported in three out of four RCTs. An improving trend in reporting concealment of patient allocation was also evident over the 15-year period. In contrast, methodological quality were more likely to have been published in a high-impact journal. Overall, the majority of published trials on POR suffer from methodological flaws and are, thus, regarded as being high-risk for bias. The same trials have used a variety of definitions for their poor responders and a variety of interventions for their head-to-head comparisons. Not surprisingly, discrepancies are also evident in the findings of trials comparing similar interventions. Based on the identified deficiencies, this novel type of 'methodology and clinical' review has introduced custom recommendations on how to improve future experimental research in the 'poor responder' population. © The Author

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  13. Evaluation of national and foreign design and application concepts for small-sized high-temperature reactors

    International Nuclear Information System (INIS)

    Hahn, L.; Nockenberg, B.

    1990-03-01

    There are currently the design types Modular HTR, MHTGR, HTTR, HTR-500, and GHR. The different design concepts are explained and are analysed with regard to their application potentials in Germany and abroad, their export into developing countries, their general and specific safety aspects, the relevant fuel cycle aspects (fuel supply and spent fuel management), and the resulting problems arising in connection with the principle of non-proliferation. In the concluding section of the report, misgivings are stated as to possible military applications of this reactor type. (DG) [de

  14. Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

    International Nuclear Information System (INIS)

    Crespi, J.C.

    1992-01-01

    This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

  15. Radiotoxicity study of a boiling water reactor core design based on a thorium-uranium fuel concept

    International Nuclear Information System (INIS)

    Nunez C, A.; Espinosa P, G.

    2007-01-01

    Full text: The innovative design of a Boiling Water Reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to 233 U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR. A comparison of the toxicity of the spent fuel showed that toxicity is lower in the thorium cycle than other commercial fuels as UO 2 and MOX (uranium and plutonium) in case of the one-through cycle for LWR. (Author)

  16. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  17. Study on concept of web-based reactor piping design data platform

    International Nuclear Information System (INIS)

    Wang Yu; Zhou Yu; Dong Jianling; Meng Yang

    2005-01-01

    For solving the piping design problems such as design data deficiency, designer communication inconvenience and design project inconsistence, Reactor Piping Design Database Platform, which is the main part of the Integrated Nuclear Project Research Platform, is proposed by analyzing the nuclear piping designs in detail. The functions and system structures of the platform are described in the paper for the sake of the realization of the Reactor Piping Design Database Platform. The platform is constituted by web-based management interface, AutoPlant selected as CAD software, and relation database management system (DBMS). (authors)

  18. Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

  19. Concept of assistance of the Federal Office for Rdiation Protection with regard to serious cases of prevention of nuclear hazards

    International Nuclear Information System (INIS)

    Becker, D.E.

    1998-01-01

    For the defence against the threats through radioactive substances, a general concept is presently being elaborated under the overall control of the Federal Government. A number of competent organisations are involved in this, for example the Federal Office of Criminal Investigation, the Federal Armed Forces, and the Federal Office for Radiation Protection. In Germany, the 16 Federal States are responsible for the prevention of nuclear hazards. In the case of hazards through radioactive material, experts from the competent radiation protection authorities are consulted. For serious cases of prevention of nuclear hazards (nuclear fuels, criticality, danger of dispersion), the Federal Office for Radiation Protection - a subordinate authority of the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety responsible for radiation protection - was given order to elaborate a concept for assistance to those Federal States. This concept is presented in the following. (author)

  20. Conceptual design of the blanket and power conversion system for a mirror hybrid fusion-fission reactor. Addendum 1. Alternate concepts. 12-month progress report addendum, July 1, 1975--June 30, 1976

    International Nuclear Information System (INIS)

    Schultz, K.R.; Dee, J.B.; Backus, G.A.; Culver, D.W.

    1976-01-01

    During the course of the Mirror Hybrid Fusion-Fission Reactor study several alternate concepts were considered for various reactor components. Several of the alternate concepts do appear to exhibit features with potential advantage for use in the mirror hybrid reactor. These are described and should possibly be investigated further in the future

  1. The Swr 1000: a nuclear power plant concept with boiling water reactor for maximum safety and economy of operation

    International Nuclear Information System (INIS)

    Brettschuh, W.

    2001-01-01

    The SWR 1000 is a design concept for a light water reactor nuclear power plant that meets all requirements regarding plant safety, economic efficiency and environ-mental friendliness. As a result of the plant's safety concept, the occurrence of core damage can, for all practical intents and purposes, be ruled out. If a core melt accident should nevertheless occur, the molten core can be retained inside the RPV, thus ensuring that all consequences of such an accident remain restricted to the plant itself. The power generating costs of the SWR 1000 are lower than with those of coal-fired and combined-cycle power plants. Power generation using nuclear energy does not release carbon dioxide to the environment, thus meeting the need for sustainable protection of our global climate. (author)

  2. A new approach for bioethanol production from sugarcane bagasse using hydrodynamic cavitation assisted-pretreatment and column reactors.

    Science.gov (United States)

    Terán Hilares, Ruly; Kamoei, Douglas Viana; Ahmed, Muhammad Ajaz; da Silva, Silvio Silvério; Han, Jong-In; Santos, Júlio César Dos

    2018-05-01

    Hydrodynamic cavitation (HC) was adopted to assist alkaline-hydrogen peroxide pretreatment of sugarcane bagasse (SCB). In the following condition: 0.29 M of NaOH, 0.78% (v/v) of H 2 O 2 , 9.95 min of process time and 3 bar of inlet pressure, 95.4% of digestibility of cellulosic fraction was achieved. To take the best use of the pretreated biomass, the overall process was intensified by way of employing a packed bed flow-through column reactor and thus enabling to handle a high solid loading of 20%, thereby leading to cellulose and hemicellulose conversions to 74.7% and 75%, respectively. In the fermentation step, a bubble column reactor was introduced to maximize ethanol production from the pretreated SCB by Scheffersomyces stipitis NRRL-Y7124, resulting in 31.50 g/L of ethanol, 0.49 g/g of ethanol yield and 0.68 g/L.h of productivity. All this showed that our HC-assisted NaOH-H 2 O 2 pretreatment strategy along with the process intensification approach might offer an option for SCB-based biorefineries. Copyright © 2018 Elsevier B.V. All rights reserved.

  3. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-01-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated

  4. An evaluation of the concept of transuranic burning using liquid metal reactors

    International Nuclear Information System (INIS)

    Rodwell, E.; Shaw, R.A.; Williams, R.F.

    1991-03-01

    The evaluation investigates the potential benefits to radioactive waste disposal of separating the transuranic elements from spent reactor fuel before the spent fuel is disposed of in geologic repositories. The evaluation addresses the question whether the benefits to radioactive waste disposal would justify processing the fuel to separate the transuranics, plus a liquid metal reactor (LMR) deployment program to transmute the separated transuranics. The evaluation concludes that adoption of a process-before-disposal policy for all the spent fuel from the light water reactors (LWRs) would accrue only modest benefits with respect to the accumulation of uranium mill tailings, the national inventory of transuranics and the licensing of a geologic repository. It is likely that this process-before-disposal policy would incur a large cost penalty, encounter major institutional difficulties, multiply licensing hurdles, and amplify political and public opposition to the overall nuclear power program. However, plutonium from spent LWR fuel is projected to be substantially more economic than enriched uranium, as fissile material for startup of LMRs when LMR deployment becomes economically justified. At that time, the LMR would fulfill its role as a reactor system that would protect the nation from diminishing energy resources. Development tasks towards defining and developing the most cost-effective LMR and associated fuel cycle remain very important. 29 refs., 3 figs., 18 tabs

  5. Modularized mirror fusion reactor concept with emphasis on fabricability, assembly, and disassembly

    International Nuclear Information System (INIS)

    Peterson, M.A.; Werner, R.W.; Hoffman, M.A.; Carlson, G.A.

    1975-01-01

    A progress report on a continuing study directed toward the development of mirror reactor designs which simultaneously satisfy the various engineering, economic, and maintenance consideration is presented. Two new blanket and coil structure designs are presented which satisfy engineering requirements equally as well as previous designs while offering substantial gains in accessibility for maintenance. Because of the commercial requirement for a high duty cycle and the possible high frequency of blanket module removal--for either maintenance replacement--the module removal must be accomplished quickly with a minimum disruption of reactor operations. The blanket and coil structure designs allow the removal of any one of the identical blanket modules without disturbing either the remaining modules or the coil and its associated support structure. With fabricated coil structure costs estimated at $2.50/lbm and the reactor net electrical power calculated from a plasma and reactor system model detailed in the paper, coil and support structure costs of between 100 to 200 $/kwe were estimated. (U.S.)

  6. Evaluation of molten fuel containment concepts for gas-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Kang, C.S.; Torri, A.

    1979-10-01

    Four in-vessel molten fuel containment concepts for the GCFR were compared, namely, (1) a ceramic crucible, (2) a borax bath, (3) a heavy metal bath, and (4) a steel bath. The ceramic crucible is the simplest but depends on substantial upward heat removal. The borax bath and the heavy metal bath concepts offer better performance but would require design changes and an increased experimental effort. The steel bath concept is a good compromise and has potential for further improvement by combining it with the essential features of other concepts, i.e., the crucible or the heavy metal bath. It is concluded that several concepts could potentially exploit the normally provided cooled liner barrier in the PCRV cavity for post-accident fuel containment

  7. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  8. Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

    Science.gov (United States)

    Cook, William Gordon

    Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

  9. Study on simulation, control and online assistance integrated system of 10 MW high temperature gas-cooled test reactor

    International Nuclear Information System (INIS)

    Luo, S.; Shi, L.; Zhu, S.

    2004-01-01

    In order to provide a convenient tool for engineering designed, safety analysis, operator training and control system design of the high temperature gas-cooled test reactor (HTR), an integrated system for simulation, control and online assistance of the HTR-10 has been designed and is still under development by the Institute of Nuclear Energy Technology (INET) of Tsinghua University in China. The whole system is based on a network environment and includes three subsystems: the simulation subsystem (SIMUSUB), the visualized control designed subsystem (VCDSUB) and the online assistance subsystem (OASUB). The SIMUSUB consists of four parts: the simulation calculating server (SCS), the main control client (MCC), the data disposal client (DDC) and the results graphic display client (RGDC), all of which can communicate with each other via network. The SIMUSUB is intended to analyze and calculate the physical processes of the reactor core, the main loop system and the stream generator, etc., as well as to simulate the normal operation and transient accidents, and the result data can be graphically displayed through the RGDC dynamically. The VCDSUB provides a platform for control system modeling where the control flow systems can be automatically generated and graphically simulated. Based on the data from the field bus, the OASUB provides some of the reactor core parameter, which are difficult to measure. This whole system can be used as an educational tool to understand the design and operational characteristics of the HTR-10, and can also provide online supports for operators in the main control room, or as a convenient powerful tool for the control system design. (authors)

  10. Expert systems for analysis and management assistance of nuclear reactor electrical power supplies

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

    Electrical power supplies of nuclear plants are very complex systems. They are studied in many ways: failure consequences, probabilistic risk assessement, failure diagnosis, corrective actions in case of incident ... Knowledge base technology (expert systems) is very suited to solve these problems. A common structural representation can generate specific functional representations; thus we get coherency and easy evolution. This paper shows a two facet methodology: the plant description (both static and dynamic) and reasoning about it. Current applications are developed for pressurized water reactors and gas cooled reactors, using the SPIRAL expert system shell buil at CEA [fr

  11. To the safety conception of the high temperature reactor with natural heat removal decay in teh case of accidents

    International Nuclear Information System (INIS)

    Petersen, K.

    1983-10-01

    On September 22, 1970, for the first time an accident simulation experiment with complete failure of the forced core cooling and the nuclear shut-down system was performed in the AVR-reactor: Due to a small heat-up of the fuel the nuclear chain-reaction was interrupted and an overheating of the core and structure was prevented due to the natural heat-convection. On the basis of the meanwhile developed computer-methods and accompanying experimental investigations it is now possible to determine exactly the behaviour of the non actively controlled core of the high temperature reactor, and to understand better the course of the AVR-experiments. On the same basis the potential and the limits of the safety conception realized in the AVR with self-stabilization in the case of accident can be determined. Such a small high temperature reactor as for example the HTR-modul of the KWU, which is characterized by a reliable and simple safety-technique with a minimum of expensive active systems, can be realized using a 2-zone-core up to a unit size of nearly 250 MW(th). (orig.) [de

  12. Three-dimensional numerical investigation of a Molten Salt reactor concept with the code CFX-5.5

    International Nuclear Information System (INIS)

    Yamaji, B.; Csom, G.; Aszodi, A.

    2002-01-01

    Partitioning and transmutation of actinides and long-lived fission products is a promising option to extend the possibilities and enhance the environmentally acceptable capabilities of nuclear energy. Also the possible implementation of the thorium cycle is considered as a way to reduce the problem of energy resources in the future. For both objectives different molten salt reactor concepts were proposed mainly based on the Molten Salt Reactor Experiment of the Oak Ridge National Laboratory. Not only critical reactors but also accelerator-driven subcritical systems (ADSs) have advantages worth considering for those aims, especially those ones with liquid fuel, such as molten salts. By using liquid fuel which is the coolant medium, too, a basically different thermalhydraulic behavior is expected than in the case of solid fuel and water coolant. In this work our purpose is to present the possible use of Computational Fluid Dynamics (CFD) technology in molten salt thermal hydraulics. The simulations were performed with the three-dimensional code CFX-5.5.(author)

  13. The materials concept in German light water reactors. A contribution to plant safety, economic performance and damage prevention

    International Nuclear Information System (INIS)

    Ilg, Ulf

    2008-01-01

    Major decisions taken as early as in the planning and construction phases of nuclear power plants may influence overall plant life. Component quality at the beginning of plant life is determined very much also by a balanced inclusion of the 'design, choice of materials, manufacturing and inspection' elements. One example of the holistic treatment of design, choice of material, and manufacture of important safety-related components in pressurized water reactors is the reactor pressure vessel (RPV) in which the ferritic compound tubes, with inside claddings, for the control rod drive nozzles are screwed into the vessel top. Also the choice of Incoloy 800 for the steam generator tubes, and the design of the main coolant pipes with inside claddings as seamless pipe bends / straight pipes with integrated nozzles connected to mixed welds with austenitic pipes are other special design features of the Siemens/KWU plants. A demonstrably high quality standard by international comparison to this day has been exhibited by the austenitic RPV internals of boiling water reactors, which were made of a low-carbon Nb-stabilized austenitic steel grade by optimum manufacturing technologies. The same material is used for backfitting austenitic pipes. Reliable and safe operation of German nuclear power plants has been demonstrated for more than 4 decades. One major element in this performance is the materials concept adopted in Germany also in the interest of damage prevention. (orig.)

  14. Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

    International Nuclear Information System (INIS)

    Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

    2017-01-01

    Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

  15. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    International Nuclear Information System (INIS)

    Sen, R. Sonat; Pope, Michael A.; Ougouag, Abderrafi M.; Pasamehmetoglu, Kemal O.

    2012-01-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities (1). Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention (2). The Deep Burn project (3) currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  16. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  17. Seasonal variation in assisted conception cycles and the influence of photoperiodism on outcome in in vitro fertilization cycles.

    Science.gov (United States)

    Wood, Simon; Quinn, Alison; Troupe, Stephen; Kingsland, Charles; Lewis-Jones, Iwan

    2006-12-01

    The effect of seasonality and daylight length on mammalian reproduction leading to spring births has been well established, and is known as photoperiodism. In assisted reproduction there is much greater uncertainty as to the effect of seasonality. This was a 4-year retrospective analysis of 2709 standardised cycles of IVF/ICSI. Data was analysed with regard to the 1642 cycles occurring during the months of extended daylight (Apr-Sept) and those 1067 cycles during winter months of restricted light length (Oct-Mar). The results showed that there was significant improvement in assisted conception outcomes in cycles performed in summer (lighter) months with more efficient ovarian stimulation 766iu v880iu/per oocyte retrieved (p=0.006). There was similarly a significantly improved implantation rate per embryo transferred 11.42% vs 9.35% (p=0.011) and greater clinical pregnancy rate 20% vs 15% (p=0.0033) during summer cycles. This study appears to demonstrate a significant benefit of increased daylight length on outcomes of IVF/ICSI cycles. Whilst the exact mechanism of this is unclear, it would seem probable that melatonin may have actions at multiple sites and on multiple levels of the reproductive tract, and may exert a more profound effect on outcomes of assisted conception cycles than has been previously considered.

  18. Scaling Studies for Advanced High Temperature Reactor Concepts, Final Technical Report: October 2014—December 2017

    Energy Technology Data Exchange (ETDEWEB)

    Woods, Brian; Gutowska, Izabela; Chiger, Howard

    2018-03-26

    Computer simulations of nuclear reactor thermal-hydraulic phenomena are often used in the design and licensing of nuclear reactor systems. In order to assess the accuracy of these computer simulations, computer codes and methods are often validated against experimental data. This experimental data must be of sufficiently high quality in order to conduct a robust validation exercise. In addition, this experimental data is generally collected at experimental facilities that are of a smaller scale than the reactor systems that are being simulated due to cost considerations. Therefore, smaller scale test facilities must be designed and constructed in such a fashion to ensure that the prototypical behavior of a particular nuclear reactor system is preserved. The work completed through this project has resulted in scaling analyses and conceptual design development for a test facility capable of collecting code validation data for the following high temperature gas reactor systems and events— 1. Passive natural circulation core cooling system, 2. pebble bed gas reactor concept, 3. General Atomics Energy Multiplier Module reactor, and 4. prismatic block design steam-water ingress event. In the event that code validation data for these systems or events is needed in the future, significant progress in the design of an appropriate integral-type test facility has already been completed as a result of this project. Where applicable, the next step would be to begin the detailed design development and material procurement. As part of this project applicable scaling analyses were completed and test facility design requirements developed. Conceptual designs were developed for the implementation of these design requirements at the Oregon State University (OSU) High Temperature Test Facility (HTTF). The original HTTF is based on a ¼-scale model of a high temperature gas reactor concept with the capability for both forced and natural circulation flow through a prismatic core with

  19. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  20. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, M.M.; Schulz, T.L.

    1993-12-07

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

  1. Nuclear reactor with makeup water assist from residual heat removal system

    Science.gov (United States)

    Corletti, Michael M.; Schulz, Terry L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  2. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Schulz, T.L.; Corletti, M.M.

    1994-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

  3. Nuclear reactor with makeup water assist from residual heat removal system

    International Nuclear Information System (INIS)

    Corletti, M.M.; Schulz, T.L.

    1993-01-01

    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

  4. Potentialities of the molten salt reactor concept for a sustainable nuclear power production based on thorium cycle in epithermal spectrum

    International Nuclear Information System (INIS)

    Nuttin, Alexis

    2002-01-01

    In the case of a significant nuclear contribution to world energy needs, the problem of present nuclear waste management pose the sustainability of the PWR fuel cycle back into question. Studies on storage and incineration of these wastes should therefore go hand in hand with studies on innovative systems dedicated to a durable nuclear energy production, as reliable, clean and safe as possible. We are here interested in the concept of molten salt reactor, whose fuel is liquid. This particularity allows an online pyrochemical reprocessing which gives the possibility to overcome some neutronic limits. In the late sixties, the MSBR (Molten Salt Breeder Reactor) project of a graphite-moderated fluoride molten salt reactor proved thus that breeding is attainable with thorium in a thermal spectrum, provided that the online reprocessing is appropriate. By means of simulation tools developed around the Monte Carlo code MCNP, we first re-evaluate the performance of a reference system, which is inspired by the MSBR project. The complete study of the pre-equilibrium transient of this 2,500 MWth reactor, started with 232 Th/ 233 U fuel, allows us to validate our reference choices. The obtained equilibrium shows an important reduction of inventories and induced radio-toxicities in comparison with the other possible fuel cycles. The online reprocessing is efficient enough to make the system breed, with a doubling time of about thirty years at equilibrium. From the reference system, we then test different options in terms of neutron economy, transmutation and control of reactivity. We find that the online reprocessing brings most of its flexibility to this system, which is particularly well adapted to power generation with thorium. The study of transition scenarios to this fuel cycle quantifies the limits of a possible deployment from the present French power stock, and finally shows that a rational management of the available plutonium would be necessary in any case. (author)

  5. An axially and radially two-zoned large liquid-metal fast breeder reactor core concept

    International Nuclear Information System (INIS)

    Kamei, T.; Arie, K.; Moriki, Y.; Suzuki, M.; Yamaoka, M.

    1985-01-01

    A new core concept that has advantages over conventional homogeneous cores in neutronics characteristics such as power peaking factor, burnup reactivity loss, and reactivity response to the movement of control rods in earthquakes has been evolved. Two options of the new core concept are feasible. One is the so-called axially heterogeneous core, with the internal blanket placed at the lower part of the core. The other concept is similar to the conventional homogeneous core, but has two different plutonium-enriched zones in the axial as well as in the radial direction, so it is a hybrid type of the conventional homogeneous core and the axially heterogeneous core. The new design concept is described and the way that the core characteristics are improved by the chosen key parameters is shown

  6. Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  7. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  8. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  9. Materials science problems of blankets in Russian concept of fusion reactor

    International Nuclear Information System (INIS)

    Solonin, M.I.

    1998-01-01

    Structural materials, beryllium and tritium breeding materials proposed for blanket of Russian reactor DEMO and Test Modules for ITER are discussed. Main requirements for the materials are concerned with basis current designs of blankets and modules and possibility meet of ones for presence and developed alloys and materials discussed considered. Main properties and results of test of ferrite-martensite and vanadium alloys for DEMO and Test Modules are cited. Beryllium compositions used as component of first wall and neutron multiplier are discussed. Liquid lithium and ceramic (lithium orthosilicate) are treated as tritium breeding materials. Russian development of reactor experimental unit for tritium breeding zone using beryllium, lithium ceramic and ferrite-martensite alloys for structural materials is presented. (orig.)

  10. The AMEBA PWR, a new concept in the technology of nuclear reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A

    2000-05-01

    AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO{sub 2}. The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up.

  11. The AMEBA PWR, a new concept in the technology of nuclear reactor safety

    International Nuclear Information System (INIS)

    Novelli, A.

    2000-01-01

    AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO 2 . The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up

  12. New reactor concepts for new generation of nuclear power in the USA: An overview

    International Nuclear Information System (INIS)

    Vujic, J.; Greenspan, E.; Milosevic, M. . E-mail addresses of corresponding authors: vujic@nuc.berkeley.edu , mmilos@vin.bg.ac.yu; Vujic, J.; Milosevic, M.)

    2005-01-01

    With the growing demands for more reliable energy sources, there is an international interest in the development of new nuclear energy systems to be deployed between 2010 and 2030, that will improve safety and reliability, decrease proliferation risks, improve radioactive waste management and lower cost of nuclear energy production. Six nuclear energy systems were selected as candidates for this Generation IV initiative. In this paper we will explore each of these concepts, as well as several of more advanced concepts. (author)

  13. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: LMFBR and HTGR advanced reactor concepts and analysis methods

    International Nuclear Information System (INIS)

    1980-01-01

    Separate abstracts are included for each of the papers presented concerning the thermal-hydraulics of LMFBR type reactors; mathematical methods in nuclear reactor thermal-hydraulics; heat transfer in gas-cooled reactors; and thermal-hydraulics of pebble-bed reactors. Two papers have been previously abstracted and input to the data base

  14. Lithium fall reactor concept: the question of jet stability, with recommendations for further experiments

    International Nuclear Information System (INIS)

    Kang, S.W.

    1978-01-01

    The stability of a liquid-lithium jet flow is of importance in a laser fusion reactor design. In this report we analyze and discuss jet stability with respect to fluid dynamics, delineating physical factors that may affect the jet breakup and performing some simple calculations to determine quantitatively the relative influences of various parameters. We define areas of uncertainty and recommend possible experimental verification, theoretical analysis, or both

  15. Hybrid reactor based on combined cavitation and ozonation: from concept to practical reality.

    Science.gov (United States)

    Gogate, P R; Mededovic-Thagard, S; McGuire, D; Chapas, G; Blackmon, J; Cathey, R

    2014-03-01

    The present work gives an in depth discussion related to the development of a hybrid advanced oxidation reactor, which can be effectively used for the treatment of various types of water. The reactor is based on the principle of intensifying degradation/disinfection using a combination of hydrodynamic cavitation, acoustic cavitation, ozone injection and electrochemical oxidation/precipitation. Theoretical studies have been presented to highlight the uniform distribution of the cavitational activity and enhanced generation of hydroxyl radicals in the cavitation zone, as well as higher turbulence in the main reactor zone. The combination of these different oxidation technologies have been shown to result in enhanced water treatment ability, which can be attributed to the enhanced generation of hydroxyl radicals, enhanced contact of ozone and contaminants, and the elimination of mass transfer resistances during electrochemical oxidation/precipitation. Compared to the use of individual approaches, the hybrid reactor is expected to intensify the treatment process by 5-20 times, depending on the application in question, which can be confirmed based on the literature illustrations. Also, the use of Ozonix® has been successfully proven while processing recycled fluids at commercial sites on over 750 oil and natural gas wells during hydraulic operations around the United States. The superiority of the hybrid process over conventional chemical treatments in terms of bacteria and scale reduction as well as increased water flowability and better chemical compatibility, which is a key requirement for oil and gas applications, has been established. Copyright © 2013 Elsevier B.V. All rights reserved.

  16. New concepts for small power reactors without on-site refuelling for non-proliferation

    International Nuclear Information System (INIS)

    Brown, N.W.; Hasberger, J.A.

    2000-01-01

    The report discusses the preliminary requirements and a technical approach for developing small reactors and the rationale for selecting them. It then discusses the four nuclear system technologies and how they might proceed to meet the requirements. Brief discussions are provided on the approaches to stimulating the appropriate international and industrial participation necessary to finance development of a design with improved proliferation resistance that is useful to the developing countries. (author)

  17. Design Concept of Kijang Research Reactor for Neutron Transmutation Doping of 300 MM ingots

    International Nuclear Information System (INIS)

    Jun, B. J.; Kim, H. S.; Seo, C. G.; Kim, H. C.; Lee, B. C.

    2013-01-01

    Neutron transmutation doping will be one of the important utilization areas of the Kijang research reactor, which is currently under design. The reactor will serve for at least 50 years. As the diameter of a current NTD ingot is already large compared to the size of the reactor, unless a provision in the reactor design is specifically made for the irradiation of potential larger diameter ingots in the future, the lifetime sustainability of the NTD activity, if possible, may be difficult to achieve. While 200 mm became the largest diameter of NTD wafers a few years ago, 300 mm is the majority nowadays in the silicon semiconductor market, and one of the world leading device companies recently invested in the construction of a 450 mm fabrication plant. The usual peak time of a wafer diameter has been around 12 years. Though the generation gap of a NTD wafer diameter has become longer as time has passed, we can foresee that NTD demand for 300 mm ingots will arise within 20 years if their NTD is possible. Our calculations show that the radial uniformity for the 300 mm ingot irradiation may be acceptable by wafer companies. However, the NTD for 450 mm ingots is judged as impractical. The KJRR is designed to irradiate 6' and 200 mm ingots to accommodate the major demands in the current and near future markets. We suppose that a 6' irradiation facility will be modified into a 300 mm irradiation facility when the demand for a 300 mm NTD arises. As the demand for the 300 mm NTD increases, other 6' and 200 mm NTD facilities will be modified one by one. A minimization of the component replacement and long-lived radwaste and a facilitation of the replacement work for each modification are important factors along with a better performance of NTD facilities

  18. Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

    International Nuclear Information System (INIS)

    Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

    2012-01-01

    Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

  19. Analysis of the transmutational characteristics of a novel molten salt reactor concept

    International Nuclear Information System (INIS)

    Csom, Gy.; Feher, S.; Szieberth, M.

    2001-01-01

    One of the arguments most frequently brought up by the opponents of the utilization of nuclear energy is the requirement that the radioactive waste and the long-lived radioisotopes accumulated in the spent fuel should be isolated for a very long time from the biosphere. The solution is the elimination of long-lived actinides (plutonium isotopes and minor actinides) and long-lived fission products by transforming (transmuting) them into short-lived or stable nuclei. The high neutron flux required for transmutation can be realized in nuclear installations. these may be conventional therma; and fast reactors, furthermore dedicated devices, namely thermal and fast reactors and accelerator driven subcritical systems (ADSs), which are specifically designed for this purpose. Some of the most promising systems are the molten salt reactors and subcritical systems, in which the fuel and material to be transmuted circulate dissolved in some molten salt. In the present paper this transmutational device, as well as recommendations for the improvement are discussed in detail (Authors)

  20. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs