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Sample records for innovative water reactor

  1. 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

  2. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    selection of chemistry controls is vital for NPPs with liquid metal cooled reactors. This paper highlights principles and approaches to chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors. The recommendations on how to arrange chemistry controls in steam/water cycles of future NPPs with innovative liquid metal cooled reactors are based taking into account: - the experience with operation of fossil power industry; - secondary side water chemistry of lead-bismuth eutectics cooled nuclear reactors at submarines; - steam/water cycles of NPPs with sodium cooled fast breeders BN-350 and BN-600; - secondary water chemistry at conventional NPPs with WER, RBMK and some other reactors. (authors)

  3. 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.)

  4. Proceedings (slides) of the OECD/NEA Workshop on Innovations in Water-cooled Reactor Technologies

    International Nuclear Information System (INIS)

    Spiler, Joze; Kim, Sang-Baik; ); Feron, Fabien; Jaervinen, Marja-Leena; Husse, Julien; ); Ferraro, Giovanni; Bertels, Frank; Denk, Wolfgang; Tuomisto, Harri; Golay, Michael; Buongiorno, J.; Todreas, N.; Adams, E.; Briccetti, A.; Jurewicz, J.; Kindfuller, V.; Srinivasan, G.; Strother, M.; Minelli, P.; Fasil, E.; Zhang, J.; Genzman, G.; Epinois, Bertrand de l'; Kim, Shin Whan; Laaksonen, Jukka; Maltsev, Mikhail; Yu, CHongxing; Powell, David; Gorgemans, Julie; Hopwood, Jerry; Bylov, Igor; Bakhmetyev, Alexander M.; Lepekhin, Andrey N.; Fadeev, Yuriy P.; Bruna, Giovanni; Gulliford, Jim; ); Ham-Su, Rosaura; Thevenot, Caroline; GAUTIER, Guy-Marie; MARSAULT, Philippe; PIGNATEL, Jean-Francois; White, Andrew; )

    2015-02-01

    New technologies and solutions have been developed over more than thirty years to improve the safety, performance and economics of nuclear power plants. Particular efforts were made in designing systems to prevent or mitigate nuclear accidents and, greatly limit or even avoid any offsite release of radioactivity. Reactor designs developed in the 1980's and later are often referred to as Generation III (Gen III) reactors. They offer enhanced safety compared to earlier Generation II (Gen II) designs, as well as improved performance and economics. Examples of Gen III safety design features include solutions for corium localisation, advanced containment structures, improved emergency core-cooling systems, filtered venting systems, hydrogen risk management solutions, etc. Some of these solutions have also been back-fitted or partially adapted to existing reactors, based on recommendations from regulators or modernisation efforts by the utilities operating these reactors, to bring their level of safety to levels approaching those of the more modern designs. Other innovations found in the latest water-cooled reactor designs include the use of passive safety systems, and often associated with those, a simplification in the design of the reactor. Gen III reactors also feature better economics, for example increased design lifetime up to 60 years, ability to use 100% MOX fuel and operate with higher flexibility, higher thermal efficiencies and reduced staff requirements. Modularity is often quoted as a feature of some Gen III designs as a way of reducing the construction times and simplifying the decommissioning of the plant. The scope of the Workshop includes, inter alia: - Evolution of regulatory and design requirements for commercial water-cooled reactors; - Innovations in water-cooled reactor technologies that allowed significant improvement in the level of safety, with a discussion on advantages and challenges of active vs. passive safety systems; - Innovations under

  5. Report on the 9th workshop on the innovative water reactor for flexible fuel cycle

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Kobayashi, Noboru; Okubo, Tsutomu; Uchikawa, Sadao

    2006-07-01

    The research on Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been performed in JAEA for the development of future innovative reactor. The workshop on FLWRs has been held every year since 1998 aiming at information exchange with other organizations such as universities, laboratories, utilities and vendors. The 9th workshop was held on March 1, 2006 under the joint auspices of JAEA and North Kanto and Kanto-Koetsu branches of Atomic Energy Society of Japan with 64 participants. The workshop began with presentation entitled 'Activities on Nuclear Science and Engineering Research and Collaboration with Industry in JAEA', followed by presentations entitled 'Progress of Research and Development on FLWR' and 'On Final Report of Feasibility Study (phase 2) on Commercialized FBR Cycle Systems'. Then two lectures followed: 'Core and Fuel Design on Super Light Water Reactor' by Tokyo University and 'Recent trends on the Development of Next Generation Nuclear Reactor' by Institute of Applied Energy. This report summarizes the lectures of the workshop. (author)

  6. Light water reactor (LWR) innovation needs in the United States: The Massachusetts Institute of Technology LWR innovation project

    International Nuclear Information System (INIS)

    Golay, M.W.

    1988-01-01

    A major effort under way within the Massachusetts Institute of Technology (MIT) Engineering School is focused on the contributions that technology innovation can make in revitalizing nuclear power in the United States. A principal component of this effort is a project to improve the designs of the next generation of light water reactors (LWRs) with emphasis on achieving improved capacity factors and safety, and reducing the construction duration. The motivation for this overall effort is to prevent the nuclear option from being unnecessarily lost by being available only in uneconomic configurations. In considering how to advance this effort, the authors focused on refining the designs of new reactors because this is the area where the greatest opportunities for improvements exist

  7. Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

    2005-01-01

    A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core 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 core in the second stage 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 core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

  8. Cogeneration of Electricity and Potable Water Using The International Reactor Innovative And Secure (IRIS) Design

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Binder, J.L.; Kostin, V.I.; Panov, Y.K.; Polunichev, V.; Ricotti, M.E.; Conti, D.; Alonso, G.

    2004-01-01

    The worldwide demand for potable water has been steadily growing and is projected to accelerate, driven by a continued population growth and industrialization of emerging countries. This growth is reflected in a recent market survey by the World Resources Institute, which shows a doubling in the installed capacity of seawater desalination plants every ten years. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh/m3 of produced desalted water. At current U.S. water use rates, a dedicated 1000 MW power plant for every one million people would be required to meet our water needs with desalted water. Nuclear energy plants are attractive for large scale desalination application. The thermal energy produced in a nuclear plant can provide both electricity and desalted water without the production of greenhouse gases. A particularly attractive option for nuclear desalination is to couple a desalination plant with an advanced, modular, passively safe reactor design. The use of small-to-medium sized nuclear power plants allows for countries with smaller electrical grid needs and infrastructure to add new electrical and water capacity in more appropriate increments and allows countries to consider siting plants at a broader number of distributed locations. To meet these needs, a modified version of the International Reactor Innovative and Secure (IRIS) nuclear power plant design has been developed for the cogeneration of electricity and desalted water. The modular, passively safe features of IRIS make it especially well adapted for this application. Furthermore, several design features of the IRIS reactor will ensure a safe and reliable source of energy and water even for countries with limited nuclear power experience and infrastructure. The IRIS-D design utilizes low-quality steam extracted from the low-pressure turbine to boil seawater in a multi-effect distillation desalination plant. The desalination plant is based on the horizontal

  9. 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

    AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads. The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany. In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented. Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of

  10. Mechanical Analysis of an Innovative Assembly Box with Honeycomb Structures Designed for a High Performance Light Water Reactor

    International Nuclear Information System (INIS)

    Herbell, Heiko; Himmel, Steffen; Schulenberg, Thomas

    2008-01-01

    The High Performance Light Water Reactor (HPLWR) is a water cooled reactor concept of the 4. generation, operated at a pressure beyond the critical point of water. Assemblies of this innovative reactor concept need to be built with assembly and moderator boxes, like boiling water reactors, to provide enough moderator water between them to compensate the low coolant density in the core. Hot, superheated steam conditions, on the other hand, require thermally insulated box walls rather than solid box walls to reduce the heat up of the moderator water. As a new an innovative approach, this paper describes moderator- and assembly boxes built from stainless steel honeycomb sandwich structures, in which the honeycomb cells are filled with alumina for thermal insulation. In comparison to solid box walls, the use of the presented design can provide the same stiffness but allows a drastic reduction of structural material and thus less neutron absorption. Finite element analyses are used to verify the required stiffness, to identify stress concentrations and to optimize the design. (authors)

  11. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  12. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  13. Investigation in justification of innovation supercritical water-cooled reactor - WWER-SCP

    International Nuclear Information System (INIS)

    Kirillov, P.L.; Baranaev, Yu.D.; Bogoslovskaya, G.P.; Glebov, A.P.; Grabezhnaya, V.A.; Kartashov, K.V.; Klushin, A.V.; Popov, V.V.

    2014-01-01

    State-of-the-art, gathered experience and development prospects of water-cooled reactors of next generation are considered. It is pointed out that development of SCWR is more attractive from the viewpoint of the basis principle of infrastructure - NPP adaptation without excessive investments. The results of experimental and calculational study of reactor installations on supercritical parameters (SCP) of water and freon are given. Consideration is given to the data on heat transfer at SCP of coolant, optimization of thermodynamic cycle, codes for thermohydraulic calculations, processes of heat and mass transfer at SCP, mass transfer and corrosion in SCP water, fuel elements and martials [ru

  14. 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

  15. Studies on the safety and transmutation behaviour of innovative fuels for light water reactors

    International Nuclear Information System (INIS)

    Schitthelm, Oliver

    2012-01-01

    Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its 238 U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

  16. 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

  17. Conceptual innovations in hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Miley, G.H.

    1980-01-01

    A number of innovations in the conception of fusion-fission hybrid reactors, including the blanket, the fusion driver, the coupling of the fusion and the fission components as well as the application of hybrid reactors are described, and their feasibility assessed

  18. 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

  19. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-09-01

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  20. Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

    Energy Technology Data Exchange (ETDEWEB)

    Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

    2017-03-15

    Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

  1. Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

    Science.gov (United States)

    Slide presentation at Conference: ASCE 7th Civil Engineering Conference in the Asian Region. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & sur...

  2. Cross cutting CFD support to innovative reactor design

    International Nuclear Information System (INIS)

    Roelofs, Ferry

    2009-01-01

    Several innovative technologies are under consideration in the world for nuclear energy production. The considered reactor systems apply either gas, sodium, lead, lead-bismuth, supercritical water, or molten salt as coolant. Therefore, methods shall be developed to determine the viability of such systems, but also to support the design of these innovative reactor systems. Computational Fluid Dynamics (CFD) is becoming more and more integrated in the daily practice of thermal-hydraulics researchers and designers. Therefore, it is very important to develop modelling approaches for the application of CFD to the specific requirements for innovative reactors. As many of these innovative reactor designs under consideration are operated using other coolants than water, one has to be careful in adopting methods which are developed for water as a coolant. Cross-cutting CFD challenges, methods and applications are presented for innovative reactors. (author)

  3. Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

    International Nuclear Information System (INIS)

    Fukaya, Yuji; Nakano, Yoshihiro; Okubo, Tsutomu

    2009-01-01

    In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal-hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.

  4. Market introduction of innovative reactors

    International Nuclear Information System (INIS)

    Heek, A.I.V.

    1996-01-01

    Besides the development of evolutionary and passive LWR, also that of innovative reactors is attractive, because other applications (new markets) besides base load electricity generation can be thought of, and interesting new features on the area of safety or waste incineration can be shown. For market introduction however, a (partial) new infrastructure and a demonstration plant are required. Taking the abundance of fossil fuels and the accompanying low fuel prices today and in the near future into account, the funds to finance this will only become available when 1)the projected energy generating costs will be substantially lower than those of today, and 2)the costs of market introduction (i.e. the demonstration plant and the required infrastructure) will be limited. Generally speaking, there are two ways to seek competitiveness of a reactor type: 1)application of economy of scale, and 2)simplification. In this paper, an example of the second possibility is pursued for an innovative reactor type. The HR1 is a 40 MWth high temperature gas cooled reactor for heat and power cogeneration, a simplified version of the German HTR Module. The power level is chosen so small that additional safety features become apparent. For example, after a total loss of coolant the fuel remains fully intact, even if the reactor shutdown system fails and the reactor goes critical again after a number of hours. These safety features are used to omit certain components, like the emergency core cooling system, or to select a cheaper version of components, e.g. replacing the containment building by a confinement. Moreover, degradation of the safety class of certain components comes within the realm of possibilities. The cost reduction offered by these two measures are used to more than offset the economy-of-scale disadvantage of this small reactor system. (author)

  5. Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems - presentation

    Science.gov (United States)

    UV disinfection is an effective process for inactivating many microbial pathogens found in source waters with the potential as stand-alone treatment or in combination with other disinfectants. For surface and groundwater sourced drinking water applications, the U.S. Environmental...

  6. Inherent safety characteristics of innovative reactors

    International Nuclear Information System (INIS)

    Heil, J.A.

    1995-11-01

    The added safety value of innovative or third generation reactor designs has been evaluated in order to determine the most suitable candidate for Dutch government funded research and development support. To this end, four innovative reactor concepts, viz. PIUS (Process Inherent Ultimate Safety), PRISM (Power Reactor Innovative Small), HTR-M (High Temperature Reactor Module) and MHTGR (Modular High Temperature Gas-cooled Reactor), have been studied and their passive and inherent safety characteristics have been outlined. Also the outlook for further technological and industrial development has been considered. The results of the study confirm the perspective of the innovative reactors for reduced dependence on active safety provisions and for a further reduced vulnerability to technical failures and human errors. The accident responses to generic accident initiators, viz. reactivity and cooling accidents, and also to reactor specific accidents show that neither active safety systems nor short term operator actions are required for maintaining the reactor core in a controlled and coolable condition. Whether this gives rise to a higher total safety of the innovative reactor designs, compared to evolutionary or advanced reactors, cannot be concluded. Supplementary experimental and analytical analyses of reactor specific accidents are required to be able to assess the safety of these innovative designs in a more quantitative manner. It is believed that the safety case of innovative reactors, which are less dependent on active safety systems, can be communicated with the general public in a more transparent way. Considering the perspective for further technological and industrial development it is not expected that any of the considered innovative reactor concepts will become commercially available within the next one to two decades. However, they could be made available earlier if they would receive sufficient financial backing. Considering the added safety perspectives

  7. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  8. The Results of Feasibility Study of Co-generation NPP With Innovative VK-300 Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    Kuznetsov, Yury N.

    2006-01-01

    The co-generation nuclear power plant (CNPP) producing electricity and district heating heat is planned to be constructed in Archangelsk Region of Russia. Following the 'Letter of Intent' signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal/h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production. (authors)

  9. The results of feasibility study of co-generation NPP with innovative VK-300 simplified boiling water reactor

    International Nuclear Information System (INIS)

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

    2004-01-01

    The co-generation nuclear power plant (CNPP) producing electricity and district-heating heat is planned to be constructed in Archangelsk Region of Russia. Following the Letter of Intent signed by Governor of Archangelsk region and by Minister of the Russian Federation for atomic energy the feasibility study of the Project has been done. The NPP will be based on the four co-generation nuclear power units with the Russian VK-300 SBWR. The innovative passive VK-300 reactor facility has been designed on the basis of well-established nuclear technologies, proven major components, the operating experience of the prototype VK-50 reactor in RIAR, Dimitrovgrad, and the experience in designing such reactors as SBWR (GE) and SWR-1000 (Siemens). The CNPP's total power is planned to be 1000 MW(e) and district-heating heat production capacity 1600 Gcal /h. A detailed description of the results of the feasibility study is presented in the report. The results of the feasibility study have shown that the Archangelsk CGNP is feasible in terms of engineering, economics and production.(author)

  10. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  11. An innovative conceptual design of the safe and simplified boiling water reactor (SSBWR) with a super-long life core

    Energy Technology Data Exchange (ETDEWEB)

    Yamashita, J. [Hitachi Ltd., Power and Industrial Systems Nuclear System Div., Ibaraki (Japan); Ohtsuka, M.; Fujimura, K.; Hidaka, M.; Nagayoshi, T. [Hitachi Ltd., Power and Industrial Systems R and D Lab., Ibaraki (Japan); Kato, Y. [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    2001-07-01

    An innovative concept for the SSBWR has been developed to provide a super-long life core of 20 years with neutron spectrum shift due to dilution from heavy to light water for coolant and to represent a passive core safety system with infinite grace period. Operability and maintainability can be largely improved by the super-long life core, reduction of the number of active components, and RPV renewal with no exchange of fuel assemblies, which can also significantly reduce the possibility of nuclear proliferation. (author)

  12. Super critical water reactors

    International Nuclear Information System (INIS)

    Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

    2005-01-01

    Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

  13. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  14. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  15. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

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

  16. International Reactor Innovative and Secure (IRIS) summary

    International Nuclear Information System (INIS)

    Carelli, Mario D.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-02-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  19. Candu reactors - experience and innovation

    International Nuclear Information System (INIS)

    Hart, R.S.; Brooks, G.L.

    1989-01-01

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  20. System and prospect assesment of the small innovative reactor IRIS-50

    International Nuclear Information System (INIS)

    Lumbanraja, Sahala M.; Wibowo

    2002-01-01

    System and prospect of the small innovative reactor IRIS-50 in Indonesia have been studied. IRIS-50 (International Reactor Innovative and Secure) is an advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse. This reactor is specifically developed to match market demands, or to developing country. This reactor is based on simplified operation and maintenance, enhanced and safety, easy to inspect, short construction time, small investment cost, competitive generating cost, and easily suited to the infrastructures. IRIS main characteristic is integral reactor concept, being all the major reactor coolant system components located inside the pressure vessel

  1. Innovative designs of nuclear reactors

    International Nuclear Information System (INIS)

    Gabaraev, B.A.; Cherepnin, Y.S.

    2010-01-01

    The world development scenarios predict at least a 2.5 time increase in the global consumption of primary energy in the first half of the twenty-first century. Much of this growth can be provided by the nuclear power which possesses important advantages over other energy technologies. However, the large deployment of nuclear sources may take place only when the new generation of reactors appears on the market and will be free of the shortcomings found in the existing nuclear power installations. The public will be more inclined to accept nuclear plants that have better economics; higher safety; more efficient management of the radioactive waste; lower risk of nuclear weapons proliferation, and provided that the focus is made on the energy option free of ∇ e 2 generation. Currently, the future of nuclear power is trusted to the technology based on fast reactors and closed fuel cycle. The latter implies reprocessing of the spent nuclear fuel of the nuclear plants and re-use of plutonium produced in power reactors

  2. Reactor water chemistry control

    International Nuclear Information System (INIS)

    Kundu, A.K.

    2010-01-01

    Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

  3. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  4. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  5. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  6. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

  7. Innovative energy production in fusion reactors

    International Nuclear Information System (INIS)

    Iiyoshi, A.; Momota, H.; Motojima, O.

    1994-01-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

  8. Innovative energy production in fusion reactors

    Science.gov (United States)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

  9. Innovative energy production in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

  10. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  11. Innovative fission reactors for this century

    International Nuclear Information System (INIS)

    Minguez, E.

    2007-01-01

    It is well known that global trends indicate a rebirth of nuclear energy due to several items: the climate change and the use of energies that emits CO 2 , the cost and dependence of gas and oil, the new innovative reactors which are competitive, safer, and sustainable and can support the Kyoto Protocol. The Advanced Reactors have safer systems than those developed in the Generation II, which demonstrates that are sustainable for the present and nuclear industry has also developed new concepts for the future which also will be sustainable. Now the new power plants that have being constructed are classified in the Generation III. Several units of this technology are in operation in Japan and other countries of the Pacific. Europe is now constructing the first unit in Finland (Olkilouto) with European technology: the European Pressurized Reactor (EPR). France has announced the beginning of the construction of an EPR in Flamanville next year. In 2000, several countries with advanced nuclear technology established the Generation IV International Forum (GIF) to develop and demonstrate nuclear energy systems that offer advantages in the following areas: sustainability, economics, safety and reliability and proliferation resistance and physical protection. These new systems will be deployed commercially after 2030. Six innovative concepts are under research, and the aim is not only produce electricity, but also hydrogen using the operational conditions of several concepts. Developed countries with NPPs in operation have strategies for the future of the nuclear energy. For the short term is to extend the operation of the NPPs until 60 years, or alternatively construction of new units of Generation III, to substitute those closed for decommissioning, keeping the percentage of contribution to the electricity generated. Between the period 2030-50, the solution is to operate the new innovative systems of the Generation IV, which uses the passive concept, and in the second part

  12. Evaluation of an Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

    Science.gov (United States)

    UV disinfection is an effective process for inactivating many microbial pathogens found in source waters with the potential as stand-alone treatment or in combination with other disinfectants. For surface and groundwater sourced drinking water applications, the U.S. Environmental...

  13. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  14. An innovative nuclear reactor as a solution to global warming

    International Nuclear Information System (INIS)

    Silva, Robson Silva da; Sefidvash, Farhang

    2007-01-01

    The problem of global warming is no longer a philosophical discussion, but it is a fact seriously threatening the future of humanity. In this paper a practical solution to the problem of global warming resulting from the fossil fuelled energy suppliers is presented. The energy conservation and alternative forms of energy such as solar, wind, and bio even though having important roles, do not satisfy the energy demand generated by an increasing world population that desires to increase its standard of living. The fission process in the nuclear reactors does not produce greenhouse gases that cause global warming. The new paradigm in nuclear energy is the future innovative reactors that meet the new standards set by the INPRO Program of the IAEA. One such a reactor is presented in this paper, namely the Fixed Bed Nuclear Reactor (FBNR) that is supported by the International Atomic Energy (IAEA) in its program of Small Reactors Without On-Site Refuelling (SRWOSR), being one of the four water cooled reactors in this program. The other three reactor concepts are PFPWR50 of Japan, BWRPB of Russia and AFPR-100 of USA. It is shown that the nuclear energy of the future is totally different than what is today in respect to safety, economics, environmental impact and proliferation. In this manner, the public perception of nuclear energy will change and its acceptability is promoted. (author)

  15. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  16. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

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

  17. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  18. Challenges and Considerations for Innovative Small and Medium Sized Reactors

    International Nuclear Information System (INIS)

    Kuznetsov, V.

    2008-01-01

    There is an ongoing interest in Member States in the development and application of small and medium sized reactors (SMRs). In the near term, most new NPPs are likely to be evolutionary water cooled reactor designs building on proven systems while incorporating technological advances and often the economics of scale, resulting from the reactor outputs of up to 1600 MW(e). For a longer term, the focus is on innovative designs aiming to provide increased benefits in the areas of safety and security, non-proliferation, waste management, resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Many innovative designs are reactors within the SMR range, having an equivalent electric power less than 700 MW(e) or even less than 300 MW(e). It is important that small or medium sized reactor does not necessarily mean small or medium sized nuclear power plant. The majority of innovative SMR concepts and designs provide for power station configurations with several units at a site or for NPP configurations with 2 or more reactor modules. In most cases, the units or modules could be added incrementally. Innovative SMRs are in many cases intended for markets different from those in which large nuclear power plants operate, i.e., markets that value more distributed electrical supplies, a better match between supply increments and investment capability or demand growth, more flexible siting or greater product variety. SMRs cannot compete with larger capacity plants on an economy of scale basis. However, they could be competitive via employing alternative design strategies, taking advantage of smaller reactor size resulting in a less complex design and operation and maintenance or in an increased overall energy conversion efficiency, and by relying on alternative deployment strategies, taking advantage of multiple unit factors and learning curve, and shorter construction schedule and 'exact' unit

  19. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  20. Qualification of FFA treatment for the water-steam cycle as an innovative lay-up strategy for the long term outage of a CANDU-6 reactor

    International Nuclear Information System (INIS)

    Ramminger, Ute; Fandrich, Jörg; Sainz, Ricardo; Ovando, Luis; Herrera, Cecilia; Mendizabal, Maribel; Dumon, Adriana; Chocron, Mauricio

    2014-01-01

    The majority of worldwide operating Nuclear Power Plants is older than 25 years, which is accompanied with extended outage duration due to large refurbishment and upgrade programs, e.g. Steam Generator Replacement and other large component replacement. For these long term outages adequate and cost effective preservation methods are required. Normally during outages, systems and components are drained and opened to atmosphere whereas wet surfaces and moisture condensation can result in uniform corrosion of carbon steel and eventually other materials; superimposed localized corrosion is possible in presence of impurities. For those systems there are in general two different lay-up methods possible. Dry lay-up by removing all water and humidity from the components or wet lay-up with demineralized and oxygen free water and additional corrosion inhibitors. Disadvantages of these lay-up methods are: High man power and hardware efforts for performing dry lay-up. Usage of hazardous chemicals like Hydrazine. Insufficient results of both lay-up methods in case of switching between dry and wet lay-up. To improve the lay-up concept for long term outages, AREVA GmbH developed an innovative concept using FFA (Film-Forming Amines) for secondary side system lay-up. The entire water-steam cycle including the Steam Generators is treated in one step without any negative impact on the treated structural materials. This technology has been applied for the first time at NPP Embalse. Embalse Nuclear Power Station consists of a CANDU-6 reactor of 648 MWe electrical output. It is in commercial operation since 1984. The shutdown for refurbishment and preparation for the second cycle of operation that includes among other tasks the replacement of the existing steam generators and power uprating has been scheduled for 2014, which causes the necessity of a lay-up optimization in the plant. This paper deals in detail with the qualification process of the FFA treatment considering the specifics

  1. European developments in single phase turbulence for innovative reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, F., E-mail: roelofs@nrg.eu [NRG, Petten (Netherlands); Rohde, M. [DUT, Delft (Netherlands); and others

    2011-07-01

    Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). They result in specific behavior of flow and heat transfer, which requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulics topics are the motivation for the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which is sponsored by the European Commission from 2010 to 2014. This paper describes the ongoing developments in an important part of this project devoted to single phase turbulence issues. To this respect, the two main issues have been identified: Non-unity Prandtl number turbulence. In case of liquid metals, molten salts or supercritical fluids, the commonly applied constant turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. This paper will report on the progress achieved with respect to the development and validation of turbulence models available in commonly used engineering tools. The paper also reports about the supporting experiments and direct numerical simulations; and, Temperature fluctuations possibly leading to thermal fatigue in innovative reactors. The status is described of a fundamental experiment dealing with the mixing of different density gases in a rectangular channel, an experiment in a more complex geometry of a small mixing plenum using a supercritical fluid, and direct numerical simulations of conjugate heat transfer on temperature fluctuations in liquid metal. (author)

  2. European developments in single phase turbulence for innovative reactors

    International Nuclear Information System (INIS)

    Roelofs, F.; Rohde, M.

    2011-01-01

    Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. From the thermal-hydraulic point of view, different innovative reactors are mainly characterized by their coolants (gas, water, liquid metals and molten salt). They result in specific behavior of flow and heat transfer, which requires specific models and advanced analysis tools. However, many common thermal-hydraulic issues are identified among various innovative nuclear systems. In Europe, such cross-cutting thermal-hydraulics topics are the motivation for the THINS (Thermal-Hydraulics of Innovative Nuclear Systems) project which is sponsored by the European Commission from 2010 to 2014. This paper describes the ongoing developments in an important part of this project devoted to single phase turbulence issues. To this respect, the two main issues have been identified: Non-unity Prandtl number turbulence. In case of liquid metals, molten salts or supercritical fluids, the commonly applied constant turbulent Prandtl number concept is not applicable and robust engineering turbulence models are needed. This paper will report on the progress achieved with respect to the development and validation of turbulence models available in commonly used engineering tools. The paper also reports about the supporting experiments and direct numerical simulations; and, Temperature fluctuations possibly leading to thermal fatigue in innovative reactors. The status is described of a fundamental experiment dealing with the mixing of different density gases in a rectangular channel, an experiment in a more complex geometry of a small mixing plenum using a supercritical fluid, and direct numerical simulations of conjugate heat transfer on temperature fluctuations in liquid metal. (author)

  3. Boiling water reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

    1975-01-01

    Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

  4. Trends on R and D of the innovative nuclear reactors in Japan

    International Nuclear Information System (INIS)

    Kinoshita, Izumi

    2002-01-01

    In Japan, since LWRs introduced from U.S.A. began their business operations one by one from 1970 and 1971, their scale-up were carried out, to reach, at present, a condition on developments of ABWR-2 of 1700 MW class in output and APWR+. They are on a line of large scale LWR development aiming at further upgrading of their economical efficiency, safety, operability and maintenance by improving and developing conventional reactors. On the other hand, an innovative small scale reactor capable of siting at proximity of its markets and flexibly responsible to needs, a low decelerated spectrum reactor intending to effectively use the resources, an super-critical pressure reactor aiming at upgrading of thermal efficiency, a high temperature gas reactor aiming at hydrogen production using nuclear heat , and so on, and so forth, are investigated at a number of institutes. And, on the fast breeder reactor, some innovative investigations such as small-scale reactor, reactor using coolant except metal sodium, and so on, in addition to development of sodium cooling large-scale reactor, under the 'Actual use strategy survey research' progressed at a center of the Japan Nuclear Cycle Development Institute, are promoted. Here were outlined on trends of R and D on various innovative reactors under classification of water cooling reactor, gas cooling reactor, and liquid metal cooling reactor. (G.K.)

  5. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

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

  6. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  7. Studies on the safety and transmutation behaviour of innovative fuels for light water reactors; Untersuchungen zum Sicherheits- und Transmutationsverhalten innovativer Brennstoffe fuer Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Schitthelm, Oliver

    2012-07-01

    Nuclear power plants contribute a substantial part to the energy demand in industry. Today the most common fuel cycle uses enriched uranium which produces plutonium due to its {sup 238}U content. With respect to the long-term waste disposal Plutonium is an issue due to its heat production and radiotoxicity. This thesis consists of three main parts. In the first part the development and validation of a new code package MCBURN for spatial high resolution burnup simulations is presented. In the second part several innovative uranium-free and plutonium-burning fuels are evaluated on assembly level. Candidates for these fuels are a thorium/plutonium fuel and an inert matrix fuel consisting of plutonium dispersed in an enriched molybdenum matrix. The performance of these fuels is evaluated against existing MOX and enriched uranium fuels considering the safety and transmutation behaviour. The evaluation contains the boron efficiency, the void coefficient, the doppler coefficient and the net balances of every radionuclide. In the third part these innovative fuels are introduced into a German KONVOI reactor core. Considering todays approved usage of MOX fuels a partial loading of one third of innovative fuels and two third of classical uranium fuels was analysed. The efficiency of the plutonium depletion is determined by the ratio of the production of higher isotopes compared to the plutonium depletion. Todays MOX-fuels transmutate about 25% to 30% into higher actinides as Americium or Curium. In uranium-free fuels this ratio is about 10% due to the lack of additional plutonium production. The analyses of the reactor core have shown that one third of MOX fuel is not capable of a net reduction of plutonium. On the other hand a partial loading with thorium/plutonium fuel incinerates about half the amount of plutonium produced by an uranium only core. If IMF is used the ratio increases to about 75%. Considering the safety behavior all fuels have shown comparable results.

  8. Next generation light water reactors

    International Nuclear Information System (INIS)

    Omoto, Akira

    1992-01-01

    In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

  9. Advanced light-water reactors

    International Nuclear Information System (INIS)

    Golay, M.W.; Todreas, N.E.

    1990-01-01

    Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

  10. Innovative hybrid biological reactors using membranes

    International Nuclear Information System (INIS)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-01-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  11. Chemistry in water reactors

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Norring, K.

    1994-01-01

    The international conference Chemistry in Water Reactors was arranged in Nice 24-27/04/1994 by the French Nuclear Energy Society. Examples of technical program areas were primary chemistry, operational experience, fundamental studies and new technology. Furthermore there were sessions about radiation field build-up, hydrogen chemistry, electro-chemistry, condensate polishing, decontamination and chemical cleaning. The conference gave the impression that there are some areas that are going to be more important than others during the next few years to come. Cladding integrity: Professor Ishigure from Japan emphasized that cladding integrity is a subject of great concern, especially with respect to waterside corrosion, deposition and release of crud. Chemistry control: The control of the iron/nickel concentration quotient seems to be not as important as previously considered. The future operation of a nuclear power plant is going to require a better control of the water chemistry than achievable today. One example of this is solubility control via regulation in BWR. Trends in USA: means an increasing use of hydrogen, minimization of SCC/IASCC, minimization of radiation fields by thorough chemistry control, guarding fuel integrity by minimization of cladding corrosion and minimization of flow assisted corrosion. Stellite replacement: The search for replacement materials will continue. Secondary side crevice chemistry: Modeling and practical studies are required to increase knowledge about the crevice chemistry and how it develops under plant operation conditions. Inhibitors: Inhibitors for IGSCC and IGA as well for the primary- (zinc) as for the secondary side (Ti) should be studied. The effects and mode of operation of the inhibitors should be documented. Chemical cleaning: of heat transfer surfaces will be an important subject. Prophylactic cleaning at regular intervals could be one mode of operation

  12. The heavy water reactors

    International Nuclear Information System (INIS)

    Brudermueller, G.

    1976-01-01

    This is a survey of the development so far of this reactor line which is in operation all over the world in various types (e.g. BHWR, PHWR). MZFR and the CANDU-type reactors are discussed in more detail. (UA) [de

  13. Nuclear reactor in deep water

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

  14. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  15. 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.

  16. Considerations in the development of safety requirements for innovative reactors: Application to modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    2003-08-01

    Member States of the IAEA have frequently requested this organization to assess, at the conceptual stage, the safety of the design of nuclear reactors that rely on a variety of technologies and are of a high degree of innovation. However, to date, for advanced and innovative reactors and for reactors with characteristics that are different from those of existing light water reactors, widely accepted design standards and rules do not exist. This TECDOC is an outcome of the efforts deployed by the IAEA to develop a general approach for assessing the safety of the design of advanced and innovative reactors, and of all reactors in general including research reactors, with characteristics that differ from those of light water reactors. This publication puts forward a method for safety assessment that is based on the well established and accepted principle of defence in depth. The need to develop a general approach for assessing the safety of the design of reactors that applies to all kinds of advanced reactors was emphasized by the request to the IAEA by South Africa to review the safety of the South African pebble bed modular reactor. This reactor, as other modular high temperature gas cooled reactors (MHTGRs), adopts very specific design features such as the use of coated particle fuel. The characteristics of the fuel deeply affect the design and the safety of the plant, thereby posing several challenges to traditional safety assessment methods and to the application of existing safety requirements that have been developed primarily for water reactors. In this TECDOC, the MHTGR has been selected as a case study to demonstrate the viability of the method proposed. The approach presented is based on an extended interpretation of the concept of defence in depth and its link with the general safety objectives and fundamental safety functions as set out in 'Safety of Nuclear Power Plants: Design', IAEA Safety Standards No. NS-R.1, issued by the IAEA in 2000. The objective

  17. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  18. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  19. Water simulation of sodium reactors

    International Nuclear Information System (INIS)

    Grewal, S.S.; Gluekler, E.L.

    1981-01-01

    The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

  20. Water Technology Innovation: 10 Market Opportunities

    Science.gov (United States)

    The Water Technology Innovation Blueprint offers an overview of market opportunities that include conserving and recovering energy, recovering nutrients, improving water infrastructure, reducing costs for water monitoring, and improving water quality.

  1. Development of next-generation light water reactor

    International Nuclear Information System (INIS)

    Ishibashi, Fumihiko; Yasuoka, Makoto

    2010-01-01

    The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

  2. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    Drevon, G.

    1983-01-01

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

  3. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  4. 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.

  5. Thorium in heavy water reactors

    International Nuclear Information System (INIS)

    Andersson, G.

    1984-12-01

    Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

  6. Reactor surface contamination stabilization. Innovative technology summary report

    International Nuclear Information System (INIS)

    1998-11-01

    Contaminated surfaces, such as the face of a nuclear reactor, need to be stabilized (fixed) to avoid airborne contamination during decontamination and decommissioning activities, and to prepare for interim safe storage. The traditional (baseline) method of fixing the contamination has been to spray a coating on the surfaces, but ensuring complete coverage over complex shapes, such as nozzles and hoses, is difficult. The Hanford Site C Reactor Technology Demonstration Group demonstrated innovative technologies to assess stabilization properties of various coatings and to achieve complete coverage of complex surfaces on the reactor face. This demonstration was conducted in two phases: the first phase consisted of a series of laboratory assessments of various stabilization coatings on metal coupons. For the second phase, coatings that passed the laboratory tests were applied to the front face of the C Reactor and evaluated. The baseline coating (Rust-Oleum No. 769) and one of the innovative technologies did not completely cover nozzle assemblies on the reactor face, the most critical of the second-phase evaluation criteria. However, one of the innovative coating systems, consisting of a base layer of foam covered by an outer layer of a polymeric film, was successful. The baseline technology would cost approximately 33% as much as the innovative technology cost of $64,000 to stabilize an entire reactor face (196 m 2 or 2116 ft 2 ) with 2,004 nozzle assemblies, but the baseline system failed to provide complete surface coverage

  7. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  8. Development Status on Innovative Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Yanagisawa, Tsutomu; Sato, Kazujiro

    2006-01-01

    The first step in Japan's nuclear fuel cycle policy is to introduce MOX recycle in light water reactors (LWRs) and the final step is to establish multiple TRU recycle in fast reactors (FRs), with the goal of realizing a stable supply, effective use of nuclear fuel resources, and the environmentally friendly production of energy. Therefore, a feasibility study on commercialized FR cycle systems has been launched since July 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC: the representative of the electric utilities) in cooperation with Central Research Institute of Electric Power Industry (CRIEPI) and vendors. In the period from July 1999 to March 2001, the feasibility study phase-I was conducted to screen out representative FR cycle concepts. In the feasibility study phase-II (April 2001 - March 2006), investigations in to the representative FR concepts were carried out to clarify the most promising concept for commercial deployment. This paper describes an innovative sodium-cooled FR, which is named as the JAEA Sodium-cooled FR (JSFR), as the most promising FR concept that meets the Generation-IV performance target. The JSFR employs several advanced technologies, such as an oxide dispersion strengthened (ODS) cladding for higher burn-up, a short-piping configuration with less elbows by adopting high chromium steel, a large scale integrated intermediate heat exchanger with a primary circulation pump, etc. Based on the design, construction and operation experiences of JOYO and MONJU, there are extensive technology bases for sodium-cooled FRs. Nevertheless, several innovative technologies implemented into the JSFR have to be developed in order to realize higher economic competitiveness by reducing construction costs and improving plant availability

  9. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  10. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  11. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

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

  12. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  13. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Presentations

    International Nuclear Information System (INIS)

    2011-01-01

    The fast reactor, which can generate electricity and breed additional fissile material for future fuel stocks, is a resource that will be needed when economic uranium supplies for the thermal reactors diminish. Further, the fast-fission fuel cycle in which material is recycled (a basic requirement to meet sustainability criteria) offers the flexibility needed to contribute decisively towards solving the problem of growing “spent” fuel inventories by greatly reducing the volume, the heat load and the radiotoxic inventory of high-level wastes that must be disposed of in long-term geological repositories. This is a waste management option that will play an increasingly important role in the future, and help to ensure that nuclear energy remains a sustainable long-term option in the world’s overall energy mix. In recognition of the fast reactor’s importance for the sustainability of the nuclear option, currently there is worldwide renewed interest in fast reactor technology development, as indicated, e.g., by the outcome of the Generation IV International Forum (GIF) technology review, which concluded with 3 out of 6 innovative systems to be fast reactors (gas cooled fast reactor, sodium cooled fast reactor, and heavy liquid metal cooled fast reactor), plus a potential fast core for a 4th concept, the super-critical water reactor. Currently, fast reactor construction projects are ongoing in India (PFBR) and Russian Federation (BN-800), whilst in China the first experimental fast reactor (CEFR) is in the commissioning phase. Fast reactor programs are also carried out in Europe (in particular in France), Japan, Republic of Korea and the USA. The most important challenges for fast reactors are in the areas of cost competitiveness with respect to LWRs and other energy sources, enhanced safety, non-proliferation, and public acceptance. With the exception of this latter, these translate into technology development challenges, i.e. the development of advanced reactor

  14. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Artioli, C.; Petrovich, Carlo; Grasso, Giacomo

    2010-01-01

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GW e y. (authors)

  15. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  16. A brief history of design studies on innovative nuclear reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2014-01-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors

  17. IRIS International Reactor Innovative and Secure Final Technical Progress Report

    International Nuclear Information System (INIS)

    Carelli, M.D.

    2003-01-01

    OAK-B135 This NERI project, originally started as the Secure Transportable Autonomous Light Water Reactor (STAR-LW) and currently known as the International Reactor Innovative and Secure (IRIS) project, had the objective of investigating a novel type of water-cooled reactor to satisfy the Generation IV goals: fuel cycle sustainability, enhanced reliability and safety, and improved economics. The research objectives over the three-year (1999-2002) program were as follows: First year: Assess various design alternatives and establish main characteristics of a point design; Second year: Perform feasibility and engineering assessment of the selected design solutions; Third year: Complete reactor design and performance evaluation, including cost assessment These objectives were fully attained and actually they served to launch IRIS as a full fledged project for eventual commercial deployment. The program did not terminate in 2002 at the end of the NERI program, and has just entered in its fifth year. This has been made possible by the IRIS project participants which have grown from the original four member, two-countries team to the current twenty members, nine countries consortium. All the consortium members work under their own funding and it is estimated that the value of their in-kind contributions over the life of the project has been of the order of $30M. Currently, approximately 100 people worldwide are involved in the project. A very important constituency of the IRIS project is the academia: 7 universities from four countries are members of the consortium and five more US universities are associated via parallel NERI programs. To date, 97 students have worked or are working on IRIS; 59 IRIS-related graduate theses have been prepared or are in preparation, and 41 of these students have already graduated with M.S. (33) or Ph.D. (8) degrees. This ''final'' report (final only as far as the NERI program is concerned) summarizes the work performed in the first four

  18. Innovative Use of Cr(VI) Plume Depictions and Pump-and-Treat Capture Analysis to Estimate Risks of Contaminant Discharge to Surface Water at Hanford Reactor Areas

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Chuck W.; Hanson, James P.; Ivarson, Kristine A.; Tonkin, M.

    2015-01-14

    The Hanford Site nuclear reactor operations required large quantities of high-quality cooling water, which was treated with chemicals including sodium dichromate dihydrate for corrosion control. Cooling water leakage, as well as intentional discharge of cooling water to ground during upset conditions, produced extensive groundwater recharge mounds consisting largely of contaminated cooling water and resulted in wide distribution of hexavalent chromium (Cr[VI]) contamination in the unconfined aquifer. The 2013 Cr(VI) groundwater plumes in the 100 Areas cover approximately 6 km2 (1500 acres), primarily in the 100-HR-3 and 100-KR-4 groundwater operable units (OUs). The Columbia River is a groundwater discharge boundary; where the plumes are adjacent to the Columbia River there remains a potential to discharge Cr(VI) to the river at concentrations above water quality criteria. The pump-and-treat systems along the River Corridor are operating with two main goals: 1) protection of the Columbia River, and 2) recovery of contaminant mass. An evaluation of the effectiveness of the pump-and-treat systems was needed to determine if the Columbia River was protected from contamination, and also to determine where additional system modifications may be needed. In response to this need, a technique for assessing the river protection was developed which takes into consideration seasonal migration of the plume and hydraulic performance of the operating well fields. Groundwater contaminant plume maps are generated across the Hanford Site on an annual basis. The assessment technique overlays the annual plume and the capture efficiency maps for the various pump and treat systems. The river protection analysis technique was prepared for use at the Hanford site and is described in detail in M.J. Tonkin, 2013. Interpolated capture frequency maps, based on mapping dynamic water level observed in observation wells and derived water levels in the vicinity of extraction and injection wells

  19. Examination of the bases for proposed innovations in reactor safety technology

    International Nuclear Information System (INIS)

    Moses, D.L.

    1986-01-01

    This paper employs the criteria for evaluations from the Nuclear Power Option Viability Study to examine the bases for proposed innovations in light water reactor safety technology. These bases for innovation fall into four broad categories as follows: (1) virtually exclusive reliance on passive safety features to preclude core damage in all situations, (2) design simplification using some passive safety features to reduce the frequency of core damage to less than about 10 -6 per reactor-year, (3) passive containment to preclude releases from any accident, and (4) designing to limit licensing attention to one or at least a few systems. Of these, only the first two, and perhaps only the second, hold significant promise for providing for the viability of advanced light water reactors

  20. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    2010-06-01

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  1. Light-water reactor research and development

    International Nuclear Information System (INIS)

    1985-05-01

    This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

  2. The safety of light water reactors

    International Nuclear Information System (INIS)

    Pershagen, B.

    1986-04-01

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  3. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  4. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  5. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    1994-03-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  6. On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho

    2007-01-01

    In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

  7. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  8. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  9. Innovative inspection system for reactor pressure vessels; Innovative Pruefsysteme fuer Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, K.; Trautmann, H.

    1999-08-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [Deutsch] Die vorgestellten kleinen, flexiblen und modernen Schwimmsysteme (DELPHIN-Manipulatoren und MIDAS-U-Boote) sind innovative Systeme fuer die Reduzierung der Aufwaende und Zeit zur Pruefung des Reaktordruckbehaelters und damit zur Reduktion der Revisionszeiten der Reaktoranlagen. (orig.)

  10. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  11. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1992-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

  12. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1993-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigates water treatment process for nuclear reactor utilization. Analysis of outwater chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic-meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agrees with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined

  13. Innovative nuclear reactor - Indian approach to meet user requirements for safety

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2002-01-01

    Full text: For sustainable development of nuclear energy, a number of key issues are to be addressed. It should be economically competitive; it must address the issues related to nuclear safety, proliferation resistance, environmental impact, waste disposal and cross cutting issues like social and infra-structural aspects. To compete successfully in the long term, in the highly competitive energy market and to overcome other challenges, it is necessary to introduce innovative reactor and fuel cycle concepts. Indian Advanced Heavy Water Reactor (AHWR) is one such innovative reactor. To guide the research and development activities related to innovative concepts, user requirements are to be formulated. User requirements covering various aspects of sustainable development are being formulated at both national and international levels. One such international project involved in the formulation of user requirements is the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). This paper deals with INPRO user requirements for safety and Indian approach to meet these requirements through AHWR

  14. Proposition of innovative and safe design of grid plate for Tehran research reactor

    International Nuclear Information System (INIS)

    Jalali, H.R.; Fadaei, A.H.

    2017-01-01

    Highlights: • An innovative and safe design for grid plate in research reactors proposed. • New grid plate acts as an independent shutdown system. • Neutronic and transient calculation was done using MTR-PC package. • Calculations show that the performance and safety of new design are acceptable. - Abstract: The purpose of this paper is to propose an innovative and safe design of grid plate for Tehran research reactor (TRR) without any reduction in its performance in comparison with the current operation. The new grid plate consisted of two joined cubic with empty walls which are place of fuels and heavy water, respectively. The proposed design is such that the reactor core is divided into two distinct parts using the heavy water. The heavy water is inserted in the walls of the new grid plate. The new design of grid plate by keeping the characteristics of the previous version creates the possibility of shutting the reactor down in critical condition. In this paper, at initial step, a simulation of acceptable benchmark for Tehran research reactor is performed which could be considered reliable and comparable with SAR (Safety Analysis Report) data. In the next step, two different designs are proposed for grid plate and then are applied to reactor core using simulation tools. For the proposed design: core excess reactivity, shutdown margin, control rod worth, neutron flux and kinetic parameters are calculated. Furthermore, the transient analysis was performed for the new design to check the status of reactor safety. Obtained results show that all neutronic parameters for the first operating core and the new design are comparable, and there is no reduction in the efficiency of reference core. Moreover, in the current design, a diverse and independent shutdown system for TRR was included. Nuclear reactor analysis codes including MTR-PC package were employed to carry out these calculations.

  15. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  16. Innovative features and fuel design approach in the iris reactor

    International Nuclear Information System (INIS)

    Petrovic, B.; Carelli, M.; Greenspan, E.; Matsumoto, H.; Padovani, E.; Ganda, F.

    2002-01-01

    The International Reactor Innovative and Secure (IRIS) is being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. The IRIS design addresses key requirements associated with advanced reactors, including improved safety, enhanced proliferation resistance, competitive electricity production cost, and improved waste management. IRIS is a modular, small/medium size (335 MWe) PWR with an integral vessel configuration. The objective has been to base its design on proven LWR technology, so that no new technology development is needed and near-term deployment is possible, yet at the same time to introduce innovative features making it attractive when compared to present PWRs. These opposing requirements resulted in an evolutionary approach to fuel and core design, balancing new features against the need to avoid extensive testing and demonstration programmes. (author)

  17. Innovative approaches to inertial confinement fusion reactors: Final report

    International Nuclear Information System (INIS)

    Bourque, R.F.; Schultz, K.R.

    1986-11-01

    Three areas of innovative approaches to inertial confinement fusion (ICF) reactor design are given. First, issues pertaining to the Cascade reactor concept are discussed. Then, several innovative concepts are presented which attempt to directly recover the blast energy from a fusion target. Finally, the Turbostar concept for direct recovery of that energy is evaluated. The Cascade issues discussed are combustion of the carbon granules in the event of air ingress, the use of alternate granule materials, and the effect of changes in carbon flow on details of the heat exchanger. Carbon combustion turns out to be a minor problem. Four ICF innovative concepts were considered: a turbine with ablating surfaces, a liquid piston system, a wave generator, and a resonating pump. In the final analysis, none show any real promise. The Turbostar concept of direct recovery is a very interesting idea and appeared technically viable. However, it shows no efficiency gain or any decrease in capital cost compared to reactors with conventional thermal conversion systems. Attempts to improve it by placing a close-in lithium sphere around the target to increase gas generation increased efficiency only slightly. It is concluded that these direct conversion techniques require thermalization of the x-ray and debris energy, and are Carnot limited. They therefore offer no advantage over existing and proposed methods of thermal energy conversion or direct electrical conversion

  18. Reactor water level measuring device

    International Nuclear Information System (INIS)

    Kuroki, Reiji; Asano, Tamotsu.

    1996-01-01

    A condensation vessel is connected to the upper portion of a reactor pressure vessel by way of a pipeline. The lower portion of the condensation vessel is connected to a low pressure side of a differential pressure transmission device by way of a reference leg pipeline. The high pressure side of the differential pressure transmission device is connected to the lower portion of the pressure vessel by way of a pipeline. The condensation vessel is equipped with a temperature sensor. When a temperature of a gas phase portion in the condensation vessel is lowered below a predetermined level, and incondensible gases in the condensation vessel starts to be dissolved in water, signals are sent from the temperature sensor to a control device and a control valve is opened. With such a constitution, CRD driving water flows into the condensation vessel, and water in which gases at the upper portion of the condensation vessel is dissolved flows into the pressure vessel by way of a pipeline. Then, gases dissolved in a reference water column in the reference leg pipeline are eliminated and the value of a reference water pressure does not change even upon abrupt lowering of pressure. (I.N.)

  19. Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, H; Nagata, A; Mingyu, Y [Tokyo Institute of Technology, Tokyo (Japan)

    2008-07-01

    A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

  20. Innovative Energy Planning and Nuclear Option Using CANDLE Reactors

    International Nuclear Information System (INIS)

    Sekimoto, H.; Nagata, A.; Mingyu, Y.

    2008-01-01

    A new reactor burn-up strategy CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) was proposed, where shapes of neutron flux, nuclide densities and power density distributions remain constant but move upward (or downward) along its core axis. This burn-up strategy can derive many merits. The change of excess reactivity along burn-up is theoretically zero for ideal equilibrium condition, and shim rods will not be required for this reactor. The reactor becomes free from accidents induced by unexpected control rods withdrawal. The core characteristics, such as power feedback coefficients and power peaking factor, are not changed during life of operation. Therefore, the operation of the reactor becomes much easier than the conventional reactors. The infinite-medium neutron multiplication factor of replacing fuel is less than unity. Therefore, the transportation and storage of replacing fuels becomes easy and safe, since they are free from criticality accidents. Small long life fast reactor with CANDLE burn-up concept has investigated with depleted uranium as a replacing fuel. Both core diameter and height are chosen to be 2.0 m, and the thermal power is 200 MW. Lead-bismuth is used as a coolant, and nitride (enriched N-15) fuel are employed. The velocity of burning region along burn-up is less than 1.0 cm/year that enables a long life design easily. The core averaged discharged fuel burn-up is about 40 percent. It is about ten times of light water reactor burn-up. The spent fuel volume becomes one-tenth of light water reactor spent fuel. If a light water reactor with a certain power output has been operated for 40 years, the CANDLE reactor can be operated for 2000 years with the same power output and with only depleted uranium left after fuel production for the light water reactor. The system does not need any reprocessing or enrichment. Therefore, the reactor operation becomes very safe, the waste

  1. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  2. Plasma engineering innovations for the ORNL TNS reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-01-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma β values up to 10 percent and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors

  3. Development of Next-Generation LWR (Light Water Reactor) in Japan

    International Nuclear Information System (INIS)

    Yamamoto, T.; Kasai, S.

    2011-01-01

    The Next-Generation Light Water Reactor development program was launched in Japan in April 2008. The primary objective of the program is to cope with the need to replace existing nuclear power plants in Japan after 2030. The reactors to be developed are also expected to be a global standard design. Several innovative features are envisioned, including a reactor core system with uranium enrichment above 5%, a seismic isolation system, the use of long-life materials and innovative water chemistry, innovative construction techniques, safety systems with the best mix of passive and active concepts, and innovative digital technologies to further enhance reactor safety, reliability, economics, etc. In the first 3 years, a plant design concept with these innovative features is established and the effectiveness of the program is reevaluated. The major part of the program will be completed in 2015. (author)

  4. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  5. High performance light water reactor

    International Nuclear Information System (INIS)

    Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

    2003-01-01

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  6. Exploring the dynamics of water innovation: Foundations for water innovation studies

    NARCIS (Netherlands)

    Wehn, U.; Montalvo, C.

    2018-01-01

    The urgent need for innovation to address multifaceted and intertwined water-related challenges is becoming increasingly clear, acknowledged and responded to with cumulating sources and amounts of funding. Nevertheless, the water sector has been claimed to be less innovative than other sectors. This

  7. An Evolutionary Approach to Water Innovation: Comparing the Water Innovation Systems in China and Europe

    DEFF Research Database (Denmark)

    Moro, Mariú Abritta

    The recent rise of the ‘green economy’ agenda has increased the attention to eco-innovations globally, with issues related to water stress identified as one of the major bottlenecks for sustainable economic growth. Water being a critical resource, more and more countries worldwide are recognizing...... the need for increasing their innovative capacity within the water sector. Using evolutionary economic theory, this thesis undertakes a longitudinal and comparative analysis of the water innovation dynamics in Europe and China, representing respectively a developed, green early mover economy......, and a centrally-planned economy and green late mover. The thesis aims to assess the similarities and differences in the mechanisms applied across these two regions, with a focus on outlining what drives eco-innovation development in the water sector. The thesis builds more specifically on the innovation system...

  8. Water level monitoring device in nuclear reactor

    International Nuclear Information System (INIS)

    Miura, Kiyohide; Otake, Tomohiro.

    1988-01-01

    Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

  9. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

    1983-01-01

    Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

  10. Development of integrated nuclear data utilization system for innovative reactors

    International Nuclear Information System (INIS)

    Naoki, Yamano; Masayuki, Igashira; Akira, Hasegawa; Kiyoshi, Kato

    2005-01-01

    An integrated nuclear data utilization system has been developing for innovative nuclear energy systems such as innovative reactors and accelerator-driven systems. The system has been constructed as a modular code system, which consists of a managing system and two subsystems. The management system named CONDUCT controls system resource management of the PC Linux server and the user authentication through Internet access. A subsystem is the nuclear data search and plotting subsystem based on a SPES engine developed by Hokkaido University. Nuclear data such as EXFOR, JENDL-3.3, ENDF/B-VI and JEFF-3.1 can be searched and plotted in the subsystem. The other is the nuclear data processing and utilization subsystem, which is able to handle JENDL-3.3, ENDF/B-VI and JEFF-3.1 to generate point-wise and group cross sections in several formats, and perform various criticality and shielding benchmarks for verification of nuclear data and validation of design methods for innovative reactors. This paper presents an overview of the integrated nuclear data utilization system, describes the progress of the system development to examine the operability of the user interface and discuss specifications of the two subsystems. (authors)

  11. Prediction of heat and mass transfer in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Ambrosini, W.; Forgione, N.; Manfredini, A.; Oriolo, F.

    2000-01-01

    This paper proposes a short review of the different forms adopted to express the analogy between heat and mass transfer for application in correlating data from condensation and evaporation experiments. In particular, the assumptions at the basis of the various forms presented by classical textbooks as well as recent research work are qualitatively discussed, proposing a unified treatment of the different models. On this background, the results of the application of one of the considered forms of the analogy to a problem having relevance for nuclear reactor safety are then discussed. The work performed in this frame is related to condensation on finned tube heat exchangers, proposed as key components in passive containment cooling systems adopted in some innovative reactor concepts. The application of the model to the experimental dana also allowed to obtain interesting information about the effect of different parameters on the cooling capabilities of this compact heat exchangers. (author)

  12. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  13. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

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

    1991-06-01

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

  14. Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

    2003-01-01

    The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

  15. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  16. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  17. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  18. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

    Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

  19. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

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

  20. Light water reactor safety research project

    International Nuclear Information System (INIS)

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

    1980-07-01

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

  1. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  2. Innovative Nuclear Reactors Implementation in the Armenian Energy Sector

    International Nuclear Information System (INIS)

    Gevorgyan, A.

    2006-01-01

    The purpose of the present paper is to demonstrate the importance of nuclear energy development in Armenia with the use of innovative nuclear reactors when considering the long-term energy planning, taking into account the specific conditions and tendencies, which are formed and developed in economy of Armenia and, in particular, in fuel-energy complex of the country. When developing the long-term program, the main factors among others considered were assumed to be the energy independence and energy security of a country, and not only the least 'cost factor', as it was usually done before. When that program was under development, such social aspects as application of the infrastructure existing within the relevant sphere, and financing of decommissioning of existing units of the Armenian NNP were also took into consideration. The studies performed have shown that implementation of innovative medium size reactors would enable the energy sector of Armenia to meet all those requirements. The issues of environmental protection were also taken into consideration when developing that program. (authors)

  3. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  4. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  6. Localized corrosion problems in water reactors

    International Nuclear Information System (INIS)

    Coriou, Henri.

    1977-01-01

    Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

  7. Status of innovative small and medium sized reactor designs 2005. Reactors with conventional refuelling schemes

    International Nuclear Information System (INIS)

    2006-03-01

    There is a renewed interest in Member States in the development and application of small and medium sized reactors (SMRs). In the near term, most new NPPs are likely to be evolutionary designs building on proven systems while incorporating technological advances and often the economics of scale, resulting from the reactor outputs of up to 1600 MW(e). For the longer term, the focus is on innovative designs aiming to provide increased benefits in the areas of safety and security, non-proliferation, waste management, resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Many innovative designs are reactors within the small-to-medium size range, having an equivalent electric power less than 700 MW(e) or even less than 300 MW(e). The projected timelines of readiness for deployment are generally between 2010 and 2030. The objective of this report is to provide Member States, including those just considering the initiation of nuclear power programmes, and those already having practical experience in nuclear power, with a balanced and objective information on important development trends and objectives of innovative SMRs for a variety of uses, on the achieved state-of-the-art in design and technology development for such reactors and on their design and regulatory status. The report is intended for many categories of stakeholders, including regulators, electricity producers, designers, non-electrical producers and policy makers. The main chapters of this report, addressed to all abovementioned groups of stakeholders, provide a summary of major specifications, applications and user-related special features of innovative SMRs, outline the achieved design and regulatory status and its progress since previous IAEA publications, review targeted deployment dates, fuel cycle options, design approaches used to meet design objectives in specific subject areas, enabling technologies and current

  8. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    International Nuclear Information System (INIS)

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E.; Lovera, P.; Fleche, J. L.; Lacroix, M.; Carra, O.; Dechelette, F.; Prele, G.; Rodriguez, G.

    2012-01-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO 2 interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

  9. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  10. Water feeding method upon reactor isolation

    International Nuclear Information System (INIS)

    Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

    1990-01-01

    The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

  11. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  12. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  13. Water chemistry in WWER reactors

    International Nuclear Information System (INIS)

    Yurmanov, V.A.; Mamet, V.A.; Shestakov, Yu.M.; Amosov, M.M.

    1997-01-01

    In this paper ''Water Chemistry in WWER Reactors'', are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH T values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ''soft decontamination'' involving changing the KOH concentration and, hence, the pH T before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs

  14. Water chemistry in WWER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V A; Mamet, V A; Shestakov, Yu M; Amosov, M M [All-Russian Scientific Research Inst. for Nuclear Power Plants Operation, Moscow (Russian Federation)

    1997-02-01

    In this paper ``Water Chemistry in WWER Reactors``, are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH{sub T} values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ``soft decontamination`` involving changing the KOH concentration and, hence, the pH{sub T} before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs.

  15. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  16. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

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

  17. Emergency cooling of presurized water reactor

    International Nuclear Information System (INIS)

    Sykora, D.

    1981-01-01

    The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

  18. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

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

  19. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    1993-07-01

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  20. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  1. Pressurised water reactor in the UK

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Since the Three Mile Island accident there has been much debate about the safety considerations of Pressurised Water Reactors. Their development will continue throughout the world but it will be based upon the lessons learned from that unfortunate accident. In the United Kingdom there is a public enquiry discussing all aspects of the reactor. The papers given in this book provide an informed addition to the literature. The design, safety and licensing and construction of a pressurised water reactor system are discussed in detail. Considerations stemming from the Three Mile Island accident are presented

  2. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  3. Water injection device for reactor container

    International Nuclear Information System (INIS)

    Sakaki, Isao.

    1996-01-01

    A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

  4. Emergency water supply facility for nuclear reactor

    International Nuclear Information System (INIS)

    Karasawa, Toru

    1998-01-01

    Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

  5. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

    International Nuclear Information System (INIS)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

  6. Nuclear desalination option for the international reactor innovative and secure (IRIS) design

    International Nuclear Information System (INIS)

    Ingersoll, D. T.; Binder, J. L.; Conti, D.; Ricotti, M. E.

    2004-01-01

    The worldwide demand for potable water is on the rise. A recent market survey by the World Resources Institute shows a doubling in desalinated water production every ten years from both seawater and brackish water sources. The production of desalinated water is energy intensive, requiring approximately 3-6 kWh per cubic meter of produced desalted water. At current U.S. water use rates, 1 kW of energy capacity per capita (or 1000 MW for every one million people) would be required to meet water needs with desalted water. The choice of the desalination technology determines the form of energy required: electrical energy for reverse osmosis systems, relatively low quality thermal energy for distillation systems, and both electrical and thermal energy for hybrid systems such as pre-heat RO systems. Nuclear energy plants are attractive for large scale desalination application. Nuclear plants can provide both electrical and thermal energy in an integrated, co-generated fashion to produce a spectrum of energy products including electricity, desalted water, process heat, district heating, and potentially hydrogen generation. A particularly attractive option for nuclear desalination is to couple it with an advanced, modular, passively safe reactor design such as the International Reactor Innovative and Secure (IRIS) plant. This allows for countries with smaller electrical grid needs and infrastructure to add new electrical and desalination capacity in smaller increments and at distributed sites. The safety by design nature of the IRIS reactor will ensure a safe and reliable source of energy even for countries with limited nuclear power experience and infrastructure. Two options for the application of the IRIS nuclear power plant to the cogeneration of electricity and desalted water are presented, including a coupling to a reverse osmosis plant and a multistage flash distillation plant. The results from an economic assessment of the two options are also presented.(author)

  7. Water desalination using different capacity reactors options

    International Nuclear Information System (INIS)

    Alonso, G.; Vargas, S.; Del Valle, E.; Ramirez, R.

    2010-01-01

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity, cogeneration of potable water production and nuclear electricity is an option to be assessed. In this paper we will perform an economical comparison for cogeneration using a big reactor, the AP1000, and a medium size reactor, the IRIS, both of them are PWR type reactors and will be coupled to the desalination plant using the same method. For this cogeneration case we will assess the best reactor option that can cover both needs using the maximum potable water production for two different desalination methods: Multistage Flash Distillation and Multi-effect Distillation. (authors)

  8. New lineup of light water reactors

    International Nuclear Information System (INIS)

    Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

    2007-01-01

    Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

  9. Supporting innovation. International Project on Innovative Nuclear Reactors and Fuel Cycles moves into first phase

    International Nuclear Information System (INIS)

    Gowin, Peter J.; Kupitz, Juergen

    2001-01-01

    Work has been initiated through the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), including technical meetings and workshops scheduled over the coming months. Among activities is an information 'side event' on INPRO at the IAEA General Conference in September 2001. Among topics addressed at the Steering Committee Meeting earlier this year are user requirements and nuclear development criteria in the area of safety; safety issues related to waste management technologies of innovative nuclear reactors and fuel cycles; methodology of assessment and comparison of innovative nuclear technology with respect to INPRO; user requirements on environmental impacts of innovative reactors, fuel cycles, and waste management; and user requirements and nuclear energy development criteria in the area of non-proliferation and proliferation resistance. In December 2001, the second meeting of the INPRO Steering Committee is scheduled. At the inaugural meeting earlier this year, the Steering Committee stressed the unique role of INPRO relative to other national and international initiatives on innovative nuclear power technologies. The role lies in identifying the needs and requirements of a spectrum of developing and developed countries; and contributing explicitly to the debate on the global acceptability of nuclear power. As of August 2001, the following countries or entities have become members of INPRO: Argentina, Canada, China, France, Germany, India, Netherlands, Russian Federation, Spain, Turkey, and the European Commission. In total, 14 experts have been nominated by their respective governments or international organizations. All IAEA Member States are also free to participate in the Steering Committee as observers. The Terms of Reference define INPRO's rationale and purpose, in the context of energy needs and developments. They state that the 'long-term outlook for nuclear energy should be considered in the broader perspective of future

  10. Simultaneous nuclear data target accuracy study for innovative fast reactors

    International Nuclear Information System (INIS)

    Aliberti, G.; Palmiotti, G.; Salvatores, M.

    2007-01-01

    The present paper summarizes the major outcomes of a study conducted within a Nuclear Energy Agency Working Party on Evaluation Cooperation (NEA WPEC) initiative aiming to investigate data needs for future innovative nuclear systems, to quantify them and to propose a strategy to meet them. Within the NEA WPEC Subgroup 26 an uncertainty assessment has been carried out using covariance data recently processed by joint efforts of several US and European Labs. In general, the uncertainty analysis shows that for the wide selection of fast reactor concepts considered, the present integral parameters uncertainties resulting from the assumed uncertainties on nuclear data are probably acceptable in the early phases of design feasibility studies. However, in the successive phase of preliminary conceptual designs and in later design phases of selected reactor and fuel cycle concepts, there will be the need for improved data and methods, in order to reduce margins, both for economic and safety reasons. It is then important to define as soon as possible priority issues, i.e. which are the nuclear data (isotope, reaction type, energy range) that need improvement, in order to quantify target accuracies and to select a strategy to meet the requirements needed (e.g. by some selected new differential measurements and by the use of integral experiments). In this context one should account for the wide range of high accuracy integral experiments already performed and available in national or, better, international data basis, in order to indicate new integral experiments that will be needed to account for new requirements due to innovative design features, and to provide the necessary full integral data base to be used for validation of the design simulation tools.

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

    International Nuclear Information System (INIS)

    Omoto, A.

    2006-01-01

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

  12. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  13. Method of measuring reactor water level

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1979-01-01

    Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

  14. International project on innovative nuclear reactors and fuel cycles

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  15. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  16. Innovations in nanotechnology for water treatment

    Directory of Open Access Journals (Sweden)

    Gehrke I

    2015-01-01

    Full Text Available Ilka Gehrke, Andreas Geiser, Annette Somborn-SchulzFraunhofer Institute for Environmental, Safety and Energy Technology UMSICHT, Oberhausen, GermanyAbstract: Important challenges in the global water situation, mainly resulting from worldwide population growth and climate change, require novel innovative water technologies in order to ensure a supply of drinking water and reduce global water pollution. Against this background, the adaptation of highly advanced nanotechnology to traditional process engineering offers new opportunities in technological developments for advanced water and wastewater technology processes. Here, an overview of recent advances in nanotechnologies for water and wastewater treatment processes is provided, including nanobased materials, such as nanoadsorbents, nanometals, nanomembranes, and photocatalysts. The beneficial properties of these materials as well as technical barriers when compared with conventional processes are reported. The state of commercialization is presented and an outlook on further research opportunities is given for each type of nanobased material and process. In addition to the promising technological enhancements, the limitations of nanotechnology for water applications, such as laws and regulations as well as potential health risks, are summarized. The legal framework according to nanoengineered materials and processes that are used for water and wastewater treatment is considered for European countries and for the USA.Keywords: nanotechnology, water technology, nanoadsorbents, nanometals, nanomembranes, photocatalysis

  17. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  18. Requirements for light water reactors

    International Nuclear Information System (INIS)

    Hedin, F.

    2009-01-01

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  19. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  20. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  1. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  2. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  3. Utility requirements for advanced light water reactors

    International Nuclear Information System (INIS)

    Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

    1996-01-01

    The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

  4. Safety aspects of pressurised water reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

  5. 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)

  6. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  7. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  8. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  9. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  10. The Consortium for Advanced Simulation of Light Water Reactors

    International Nuclear Information System (INIS)

    Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul

    2011-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  11. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

  12. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Sawa, Toshio; Takahashi, Sankichi; Takashima, Yoshie.

    1983-01-01

    Purpose: To efficiently eliminate radioactive materials such as iron oxide and cobalt ions with less heat loss by the use of an electrode assembly applied with a direct current. Constitution: In a reactor water clean-up device adapted to pass reactor water through an electrode assembly comprising at least a pair of anode and cathode applied with a direct current to eliminate various types of ions contained in the reactor water by way of the electrolysis or charge neutralization at the anode, the cathode is constituted with a corrosion resistant grid-like or porous metal plate and a layer to the upper portion of the metal plate filled with a plurality of metal spheres of about 1 - 5 mm diameter, and the anode is made of insoluble porous or spirally wound metal material. (Seki, T.)

  13. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  14. Safety of evolutionary and innovative nuclear reactors: IAEA activities and world efforts

    International Nuclear Information System (INIS)

    Saito, T.; Gasparini, M.

    2004-01-01

    'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  16. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

    2003-01-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  17. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2009-01-01

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  18. 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)

  19. Beyond the light water reactor

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1980-01-01

    One of the strong interests in examining alternative nuclear fuel cycles is to search for schemes that are more efficient than LWRs in their use of uranium, but that do not carry the additional proliferation hazard associated with widespread plutonium utilization. One possibility is to improve the uranium efficiency of current reactor types by other means than recycling. A second possibility, offering greater potential for improvement, is to utilize thorium-uranium fuel cycles in which uranium-233 is denatured by the addition of uranium-238, making enrichment necessary to yield weapons-usable material. The bulk of the reactor's fuel material would be thorium-232, so that most of the fissile material produced would be uranium-233. It is important to recognize that these two possibilities - once through improvements and denatured thorium-uranium - could be introduced sequentially in reactor types that are currently in use. Fuel cycles may change over time, but it is equally significant from the point of view of non-proliferation that they will also vary from place to place and, most importantly, from country to country. The author argues that alternative nuclear power systems and a slower growth may affect the diversion of nuclear materials to weapons. A real question, though, is whether we have time to explore the possibilities. It has become apparent that predictions made of the growth rate for nuclear power were too high. The 1000 large power plants the US was to have by the year 2000 have been reduced to fewer than 300. This reduces the pressure, resulting from our limited uranium resources, to push the LMFBR. Extra time gives us a chance to examine the possibilities

  20. EPRI program in water reactor safety

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

    1975-01-01

    The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

  1. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  2. The water chemistry of CANDU PHW reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. Study of the economic viability of the innovative nuclear reactor SMART in Brazil

    International Nuclear Information System (INIS)

    Escanhoela, Cordelia Mara Fazzio; Lima, Ana Cecilia de Souza; Sabundjian, Gaianê

    2017-01-01

    The main objective of this study is to evaluate the economic viability of the installation and operation of the innovative System - Integrated Modular Advanced Reactor (SMART) in Brazil. SMART, developed by the Korea Atomic Energy Research Institute (KAERI), is a small and modular Power Water Reactor (PWR), presents electric power of 100 MW and thermal power of 330 MW; it has a passive safety system and integral refrigeration configuration, characteristics that, allied with modularization, simplification and technological improvements, give SMART greater reliability and economy when compared to conventional reactors. SMART presents, in addition to electricity production, the functions of seawater desalination and district heat generation. The research is based on projections of energy demand in the medium and long term with emphasis on electricity and search for the reduction of greenhouse gases. These previsions indicate the need for energy expansion and diversification of the current sources in Brazil, predominantly water sources. The methodology used is based on the cost of electric generation, production capacity and construction time of SMART, adopting the investment model similar to the Angra 3 plant and the use of mirrored costs between the plants. The feasibility of the project was evaluated through the financial criteria: Internal Rate of Return (IRR), Net Present Value (NPV) and Weighted Average Cost of Capital (WACC), whose revenue should be generated through a tariff passed on to the consumer. (author)

  5. Study of the economic viability of the innovative nuclear reactor SMART in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Escanhoela, Cordelia Mara Fazzio; Lima, Ana Cecilia de Souza; Sabundjian, Gaianê, E-mail: liafazzio@hotmail.com, E-mail: aclima@ipen.br, E-mail: gdjian@ipen.br [Centro de Engenharia Nuclear - CEN, Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The main objective of this study is to evaluate the economic viability of the installation and operation of the innovative System - Integrated Modular Advanced Reactor (SMART) in Brazil. SMART, developed by the Korea Atomic Energy Research Institute (KAERI), is a small and modular Power Water Reactor (PWR), presents electric power of 100 MW and thermal power of 330 MW; it has a passive safety system and integral refrigeration configuration, characteristics that, allied with modularization, simplification and technological improvements, give SMART greater reliability and economy when compared to conventional reactors. SMART presents, in addition to electricity production, the functions of seawater desalination and district heat generation. The research is based on projections of energy demand in the medium and long term with emphasis on electricity and search for the reduction of greenhouse gases. These previsions indicate the need for energy expansion and diversification of the current sources in Brazil, predominantly water sources. The methodology used is based on the cost of electric generation, production capacity and construction time of SMART, adopting the investment model similar to the Angra 3 plant and the use of mirrored costs between the plants. The feasibility of the project was evaluated through the financial criteria: Internal Rate of Return (IRR), Net Present Value (NPV) and Weighted Average Cost of Capital (WACC), whose revenue should be generated through a tariff passed on to the consumer. (author)

  6. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    1988-12-01

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  7. 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

  8. Historical perspective of thermal reactor safety in light water reactors

    International Nuclear Information System (INIS)

    Levy, S.

    1986-01-01

    A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

  9. Innovative research reactor core designed. Estimation and analysis of gamma heating distribution

    International Nuclear Information System (INIS)

    Setiyanto

    2014-01-01

    The Gamma heating value is an important factor needed for safety analysis of each experiments that will be realized on research reactor core. Gamma heat is internal heat source occurs in each irradiation facilities or any material irradiated in reactor core. This value should be determined correctly because of the safety related problems. The gamma heating value is in general depend on. reactor core characteristics, different one and other, and then each new reactor design should be completed by gamma heating data. The Innovative Research Reactor is one of the new reactor design that should be completed with any safety data, including the gamma heating value. For this reasons, calculation and analysis of gamma heating in the hole of reactor core and irradiation facilities in reflector had been done by using of modified and validated Gamset computer code. The result shown that gamma heating value of 11.75 W/g is the highest value at the center of reactor core, higher than gamma heating value of RSG-GAS. However, placement of all irradiation facilities in reflector show that safety characteristics for irradiation facilities of innovative research reactor more better than RSG-GAS reactor. Regarding the results obtained, and based on placement of irradiation facilities in reflector, can be concluded that innovative research reactor more safe for any irradiation used. (author)

  10. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  11. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-12-01

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  12. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

    1992-01-01

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  13. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    The Quality Assurance Program used by Westinghouse Nuclear Energy Systems Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements.

  14. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  15. Facilitation of decommissioning light water reactors

    International Nuclear Information System (INIS)

    Moore, E.B. Jr.

    1979-12-01

    Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

  16. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  17. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  18. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  19. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  20. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  1. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  2. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  3. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  4. Advanced and sustainable fuel cycles for innovative reactor systems

    International Nuclear Information System (INIS)

    Glatz, J. P.; Malmbeck, R.; Purroy, D. S.; Soucek, P.; Inoue, T.; Uozumi, K.

    2007-01-01

    reprocessing facility with improved economics and the higher radiation stability of the molten salt media are some of the arguments in favour of pyro-reprocessing. Adaptations of this technology exist for the treatment of both oxide and nitride fuels. The flowsheet for the treatment of nitride fuels is similar to that of metal fuel. In the case of oxides a head-end reduction step is needed. It can be performed by direct electroreduction, where the heat generating fission products are removed and the fissile materials are recovered as an alloy, which can be again directly reprocessed by electrorefining. The present paper describes the progress made at ITU - mainly in the frame of the network projects mentioned above - in developing the grouped actinide recycling concept with a view to the sustainability goals described above for innovative reactor systems. In the frame of these projects, reprocessing of EBRII type metallic alloy fuel with 2% of Am and 5% of lanthanides (U 6 0Pu 2 0-Zr 1 0Am 2 Nd 3 .5Y 0 .5Ce 0 .5Gd 0 .5) is being carried out by electrorefining at ITU. An excellent grouped separation of actinides from lanthanides (An/Ln mass ratio = 2400) had been obtained. The high neutron capture of lanthanides and their possibly detrimental interaction with the cladding material implies that they must be separated. In this sense the choice of the cathode material for the actinide recovery is essential and it could be shown that aluminium is an excellent material for a pyrochemical partitioning process. The results are confirmed in conditions simulating the scaling up (multiple run) of the process, with an accumulation of Ln in the salt. One of the major goals is the minimization of actinide losses and to thereby reduce significantly the radiotoxicity of the waste produced. The results shown here represent the first demonstration of an efficient grouped actinide recovery from realistic metallic fuels and are therefore an important step in achieving the sustainability goals of

  5. Trends in light water reactor dosimetry programs

    International Nuclear Information System (INIS)

    Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

    1977-01-01

    Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

  6. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  7. 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

  8. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    International Nuclear Information System (INIS)

    Pahladsingh, R.R.

    2002-01-01

    Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

  9. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

    2006-01-01

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

  10. 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

  11. Towards intrinsically safe light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  12. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  13. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  14. Performance of an innovative multi-stage anaerobic reactor during ...

    African Journals Online (AJOL)

    Start-up of an anaerobic reactor is a relatively delicate process and depends on various factors such as wastewater composition, available inoculum, operating conditions and reactor configuration. Accordingly, systematized operational procedures are important, mainly during the start-up of an anaerobic reactor.

  15. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    Giedraityte, Zivile

    2008-01-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

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

    International Nuclear Information System (INIS)

    Mourogov, V.M.

    2001-01-01

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

  17. Medium-sized water reactors for undeveloped regions

    International Nuclear Information System (INIS)

    Osmachkin, V. S.

    2004-01-01

    In the new century the growth of population and an increasing of energy demands together with the difficulties of fossil fuel supply are expected. It is important to find optimal ways in solving such problems without the climate warming. The nuclear power having many advantages in comparison with fossil fuel technologies could play the great role in near future. The Medium-Sized Nuclear Reactors for production of electricity, heat and fresh water are considered as a main direction of nuclear power applications in the developing world It is important to discuss the requirements to such nuclear plants for using in the Countries with Small and Medium Electricity Grids. Particularly, cost-benefit analysis of construction NPP has to include assessment of all type risks and effectiveness of plant. In the paper an attention is paid on Water Reactors designed on the basis of navy technology. Such compact PWR built on special mills and placed on special floating vessel could be used in undeveloped regions. Total plant can be transported to any point of World Ocean and return back to mill for repair or decommissioning after exhaustion of lifetime. It is expected that such reactors with innovative design approach, provision of high safety and proper economic efficiency, based on leasing procedures, could be very attractive for medium-sized and developing countries.(author)

  18. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  19. Integrated infrastructure initiatives for material testing reactor innovations

    International Nuclear Information System (INIS)

    Dekeyser, Jean; Vermeeren, Ludo; Iracane, Daniel

    2011-01-01

    Highlights: → The EU FP7 MTR+I3 project has initiated a durable cooperation between MTR operators. → Improvements in irradiation test device technology and instrumentation were achieved. → Professional training efforts were streamlined and best practices were exchanged. → A framework has been set up to coordinate and optimize the use of MTRs in the EU. - Abstract: The key goal of the European FP6 project MTR+I3 was to build a durable cooperation between Material Testing Reactor (MTR) operators and relevant laboratories that can maintain European leadership with updated capabilities and competences regarding reactor performances and irradiation technology. The MTR+I3 consortium was composed of 18 partners with a high level of expertise in irradiation-related services for all types of nuclear plants. This project covered activities that foster integration of the MTR community involved in designing, fabricating and operating irradiation devices through information exchange, know-how cross-fertilization, exchanges of interdisciplinary personnel, structuring of key-technology suppliers and professional training. The network produced best practice guidelines for selected irradiation activities. This project allowed to launch or to improve technical studies in various domains dealing with irradiation test device technology, experimental loop designs and instrumentation. Major results are illustrated in this paper. These concern in particular: on-line fuel power determination, neutron screen optimization, simulation of transmutation process, power transient systems, water chemistry and stress corrosion cracking, fission gas measurement, irradiation behaviour of electronic modules, mechanical loading under irradiation, high temperature gas loop technology, heavy liquid metal loop development and safety test instrumentation. One of the major benefits of this project is that, starting from a situation of fragmented resources in a strongly competitive sector, it has

  20. What drives innovation in nuclear reactors technologies? An empirical study based on patent counts

    International Nuclear Information System (INIS)

    Berthelemy, Michel

    2012-01-01

    This paper examines the evolution of innovation in nuclear power reactors between 1974 and 2008 in twelve OECD countries and assesses to what extent nuclear innovation has been driven by economic incentives, political decisions and safety regulation considerations. We use priority patent applications related to Nuclear Power Plants (NPPs) as a proxy for innovating activity. Our results highlight that nuclear innovation is partly driven by the conventional paradigm where both demand-pull, measured by NPPs constructions in the innovating country and in the rest of the world, and technology-push, measured by Research and Development (R and D) expenditures specific to NPPs, have a positive and significant impact on innovation. Our results also evidence that the impact of public R and D expenditures and national NPPs construction on innovation is stronger when the quality of innovation, measured by forward patent citations, is taken into account, and have a long run positive impact on innovation through the stock of knowledge available to innovators. In contrast, we show that political decisions following the Three Miles Island and Chernobyl nuclear accidents, measured by NPPs cancellations, have a negative impact on nuclear innovation. Finally, we find that the nuclear safety authority has an ambivalent effect on innovation. On one hand, regulatory inspections have a positive impact on innovation, one the other hand, regulatory decisions to temporarily close a NPP have an adverse impact on innovation. (author)

  1. Heavy water upgrading system in the Fugen heavy water reactor

    International Nuclear Information System (INIS)

    Matsushita, T.; Susaki, S.

    1980-01-01

    The heavy water upgrading system, which is installed in the Fugen heavy water reactor (HWR) was designed to reuse degraded heavy water generated from the deuteration-dedeuteration of resin in the ion exchange column of the moderator purification system. The electrolysis method has been applied in this system on the basis of the predicted generation rate and concentration of degraded heavy water. The structural feature of the electrolytic cell is that it consists of dual cylindrical electrodes, instead of a diaphragm as in the case of conventional water electrolysis. 2 refs

  2. The IAEA's international project on innovative nuclear reactors and fuel cycles (INPRO)

    International Nuclear Information System (INIS)

    Kuptiz, Juergen; )

    2002-01-01

    This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited all interested Member states to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology and invited Member states to consider to contribute to a task force on innovative nuclear reactors and fuel cycle

  3. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  4. The chemistry of water reactor fuel

    International Nuclear Information System (INIS)

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  5. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  6. Materials technologies of light water reactors

    International Nuclear Information System (INIS)

    Begley, R.

    1984-01-01

    Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

  7. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  8. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

  9. Advanced light water reactors for the nineties

    International Nuclear Information System (INIS)

    Ross, F.A.; Sugnet, W.R.

    1987-01-01

    The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

  10. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

    In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

  11. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  12. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  13. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  14. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  15. Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

    Energy Technology Data Exchange (ETDEWEB)

    Simon, N.; Lorcet, H.; Beauchamp, F.; Guigues, E. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Lovera, P.; Fleche, J. L. [CEA, DEN, DPC Saclay, F-91191 Gif-sur-Yvette (France); Lacroix, M. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Carra, O. [AREVA / NP, 10 Rue Juliette Recamier, 69003 Lyon (France); Dechelette, F. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France); Prele, G. [EDF/SEPTEN, 12-14 avenue Dutrievoz, 69628 Villeurbane Cedex (France); Rodriguez, G. [CEA, DEN, DTN Cadarache, F-13108 Saint-Paul-lez-Durance (France)

    2012-07-01

    Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, a functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  18. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-02-01

    The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

  19. Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

    Energy Technology Data Exchange (ETDEWEB)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-07-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

  20. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  1. 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)

  2. 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)

  3. Neutron disadvantage factors in heavy water and light water reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  4. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  5. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  6. Study of Some Innovant Reactors without on- Site Refueling with Triso and Cermet Fuel

    OpenAIRE

    A.Chetaine; A. Benchrif; H. Amsil; V. Kuznetsov; Y. Shimazu

    2012-01-01

    The evaluation of unit cell neutronic parameters and lifetime for some innovant reactors without on sit-refuling will be held in this work. the behavior of some small and medium reactors without on site refueling with triso and cermet fuel. For the FBNR long life except we propose to change the enrichment of the Cermet MFE to 9%. For the AFPR reactor we can see that the use of the Cermet MFE can extend the life of this reactor but to maintain the same life period for AFPR...

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  8. Economics of seawater desalination with innovative nuclear reactors and other energy sources: the EURODESAL project

    International Nuclear Information System (INIS)

    Nisan, S.; Volpi, L.

    2004-01-01

    This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, which were carried out by a consortium of four EU and one Canadian, Industrials and two leading EU R and D organisations. Major results of the project, in particular of its economic evaluation work package as discussed in this paper, are: 1. A coherent demonstration of the technical feasibility of nuclear desalination through the development of technical principles for the optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies; verification that the integrated system design does not adversely affect nuclear reactor safety. 2. The development of codes and methods for an objective assessment of the competitiveness and sustainability of proposed solutions through comparison, in European conditions, with fossil and renewable energy based solutions. The results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options. Thus, for example, specific desalination costs ($/m 3 of desalted water) for nuclear systems such as the AP600 and the French PWR900 (reference base case), coupled to Multiple Effect Distillation (MED) or the Reverse Osmosis (RO) processes, are 30% to 60% lower than fossil energy based systems using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rates = 10%, low fossil fuel prices) desalination costs with the nuclear options with the nuclear reactors are 7% to 15% lower, depending upon the desalination capacities. Furthermore, with the high performance coupling schemes developed by the EURODESAL partners, the specific desalination costs of nuclear systems are reduced by another 2% to 14%, even without system and design optimisation. (author)

  9. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  10. Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

    International Nuclear Information System (INIS)

    Gu Yingbin

    2012-01-01

    Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

  11. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    Cullingford, H.S.; Edeskuty, F.J.

    1981-01-01

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  12. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  13. Neutronics conceptual design of the innovative research reactor core using uranium molybdenum fuel

    International Nuclear Information System (INIS)

    Tukiran S; Surian Pinem; Tagor MS; Lily S; Jati Susilo

    2012-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest. Reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR) and then as an alternative to replace the old research reactor. The aim of this research is to get the optimal configuration of equilibrium core with the acceptance criteria are minimum thermal neutron flux is 2.5E14 n/cm 2 s at the power level of 20 MW (minimum), length of cycle of more than 40 days, and the most efficient of using fuel in the core. Neutronics design has been performed for new fuel of U-9Mo-AI with various fuel density and reflector. Design calculation has been performed using WIMSD-5B and BATAN-FUEL computer codes. The calculation result of the conceptual design shows four core configurations namely 5x5, 5x7, 6x5 and 6x6. The optimalization result for equilibrium core of innovative research reactor is the 5x5 configuration with 450 gU fuel loading, berilium reflector, maximum thermal neutron flux at reflector is 3.33E14 n/cm 2 sand length of cycle is 57 days is the most optimal of IRR. (author)

  14. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  15. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  16. Nuclear energy. The innovations of the N4 reactor

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

    The coupling to the electric network of the two first units of N4 type reactors, on the site of Chooz in the Ardennes, marks the third great step of the French nuclear programme of PWR type reactors, after the realization of 34 units of 900 MWe and 20 units of 1300 M We. The nuclear boiler N4, realizes a new evolution in power, in performances and in reliability. (N.C.)

  17. User requirements in the area of safety of innovative nuclear reactors and fuel cycle installations

    International Nuclear Information System (INIS)

    Kuczera, B.; Juhn, P.E.; Fukuda, K.; )

    2002-01-01

    Full text: Against the background of already existing IAEA and INSAC publications in the area of safety, in the framework of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a set of user requirements for the safety of future nuclear installations has been established. Five top-level requirements are expected to apply to any type of innovative design. They should foster an increased level of safety that is transparent to and fully accepted by the general public. The approach to future reactor safety includes two complementary strategies: increased emphasis on inherent safety characteristics and enhancement of defense in depth. As compared to existing plants, the effectiveness of preventing measures should be highly enhanced, resulting in fewer mitigation measures. The targets and possible approaches of each of the five levels of defense developed for innovative reactor designs are outlined in the paper

  18. Control of water chemistry in operating reactors

    International Nuclear Information System (INIS)

    Riess, R.

    1997-01-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ''modified'' B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs

  19. Control of water chemistry in operating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riess, R [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ``modified`` B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs.

  20. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  1. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  2. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  3. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

    Rothwell, G.; Van der Zwaan, B.

    2001-01-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  4. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  5. Westinghouse Water Reactor Divisions quality assurance plan

    International Nuclear Information System (INIS)

    1977-09-01

    The Quality Assurance Program used by Westinghouse Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements. This program satisfies the NRC Quality Assurance Criteria, 10CFR50 Appendix B, to the extent that these criteria apply to safety related NSSS equipment. Also, it follows the regulatory position provided in NRC regulatory guides and the requirements of ANSI Standard N45.2.12 as identified in this Topical Report

  6. Instrument lance for boiling water reactors

    International Nuclear Information System (INIS)

    Proell, N.; Bertz, S.; Graebener, K.H.

    1980-01-01

    The instrument lance contains in the lance cover pipe a thimble as part of the drive chamber system. Other thimbles serve to carry neutron detectors. Detectors can be exchanged without opening the reactor pressure vessel and without removing the fuel elements. Furthermore the detector exchange is independent from the fuel element cycle. The measurement lance passes from the bottom of the pressure vessel over the total hight of the core in the water ducts between the fuel elements and can thus determine the neutron flux distribution. (DG) [de

  7. Light-water-reactor hydrogen manual

    International Nuclear Information System (INIS)

    Camp, A.L.; Cummings, J.C.; Sherman, M.P.; Kupiec, C.F.; Healy, R.J.; Caplan, J.S.; Sandhop, J.R.; Saunders, J.H.

    1983-06-01

    A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures

  8. Development trends in light water reactors

    International Nuclear Information System (INIS)

    Fogelstroem, L.; Simon, M.

    1988-01-01

    The present market for new nuclear power plants is weak, but is expected to pick up again, which is why great efforts are being made to further develop the light water reactor line for future applications. There is both a potential and a need for further improvement, for instance with respect to even higher cost efficiency, a simplified operating permit procedure, shorter construction periods, and increased operational flexibility to meet rising demands in load following behavior and in better cycle data of fuel elements. However, also public acceptance must not be forgotten when deciding about the line to be followed in the development of LWR technology. (orig.) [de

  9. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  10. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  11. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  12. Thermodynamic analysis of a supercritical water reactor

    International Nuclear Information System (INIS)

    Edwards, M.

    2007-01-01

    A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

  13. Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan

    2002-01-01

    There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international

  14. IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2011-01-01

    The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

  15. IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Working Material

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of the TM was to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus was on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

  16. IAEA Technical Meeting on Innovative Fast Reactor Designs with Enhanced Negative Reactivity Feedback Features. Presentations

    International Nuclear Information System (INIS)

    2012-01-01

    The objective of the TM is to review and discuss the safety characteristics and the performances of the core of innovative fast reactor concepts, as well as to present the ongoing R&D activities in the area of core design and advanced simulation tools and methods for fast reactor core physics analysis. The focus is on fast spectrum cores optimized for actinide utilization and transmutation and, in particular, on core designs with enhanced negative reactivity feedback effects

  17. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  18. 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.

  19. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  20. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

    2016-01-01

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  1. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  2. 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

  3. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  4. Carbon-14 in reactor plant water

    International Nuclear Information System (INIS)

    Knowles, G.K.

    1979-01-01

    The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

  5. Summary of the 4th workshop on the reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  6. Summary of the 4th workshop on the reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

    2001-09-01

    The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

  7. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

  8. Light a CANDLE. An innovative burnup strategy of nuclear reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi

    2005-11-01

    CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)

  9. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    Bouchard, J.; Kanevoky, A.; Reuss, P.

    1975-01-01

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

  10. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  11. IRIS - Generation IV Advanced Light Water Reactor for Countries with Small and Medium Electricity Grids

    International Nuclear Information System (INIS)

    Carelli, M. D.

    2002-01-01

    An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a Generation IV Reactor, International Reactor Innovative and Secure (IRIS). IRIS is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., fuel cycle sustainability, enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it does not require new technology development since it relies on the proven technology of light water reactors. This paper presents the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and four-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. The path forward for possible future extension to a eight-year cycle will be also discussed. IRIS has a large potential worldwide market because of its proven technology, modularity, low financing, compatibility with existing grids and very limited infrastructure requirements. It is especially appealing to developing countries because of ease of operation and because its medium power is more adaptable to smaller grids. (author)

  12. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

    Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

  13. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  14. Development of supercritical water reactors in Russia and abroad

    International Nuclear Information System (INIS)

    Glebov, A.P.; Klushin, A.V.

    2014-01-01

    The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

  15. International R and D project on development of coated particle fuel for innovative reactors

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    The paper presents an outline for an international collaborative project of coated particle fuel development for innovative reactors. Specific issues include identification of R and D needs and the Member State facilities for meeting the needs followed by development and demonstration of technology. (author)

  16. EURATOMM RDDD in innovative reactor systems and fuel cycles

    International Nuclear Information System (INIS)

    Van Goethem, G.

    2008-01-01

    In this introductory overview paper, the following questions are addressed: 1) What are the innovation challenges in energy technologies, in particular, in nuclear fission, that the European industry and research organisations are faced with? In the short (today), medium (2015) and long term (2040)? 2) What kind of response do the EURATOM research, development, demonstration and deployment (RDDD) programmes offer as solution to the above challenges? And what has been achieved so far, using the Community instruments? (author)

  17. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  18. Nuclear fuel for light water reactors

    International Nuclear Information System (INIS)

    Etemad, A.

    1976-01-01

    The goal of the present speech is to point out some of the now-a-day existing problems related to the fuel cycle of light water reactors and to foresee their present and future solutions. Economical aspects of nuclear power generation have been considerably improving, partly through technological advancements and partly due to the enlargement of unit capacity. The fuel cycle, defined in the course of this talk, discusses the exploration, mining, ore concentration, purification, conversion, enrichment, manufacturing of fuel elements, their utilization in a reactor, their discharge and subsequent storage, reprocessing, and their re-use or disposal. Uranium market in the world and the general policy of several uranium owning countries are described. The western world requirement for uranium until the year 2000, uranium resources and the nuclear power programs in the United States, Australia, Canada, South Africa, France, India, Spain, and Argentina are discussed. The participation of Iran in a large uranium enrichment plant based on French diffusion technology is mentioned

  19. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

  20. Hydrogen behavior in light-water reactors

    International Nuclear Information System (INIS)

    Berman, M.; Cummings, J.C.

    1984-01-01

    The Three Mile Island accident resulted in the generation of an estimated 150 to 600 kg of hydrogen, some of which burned inside the containment building, causing a transient pressure rise of roughly 200 kPa (2 atm). With this accident as the immediate impetus and the improved safety of reactors as the long-term goal, the nuclear industry and the Nuclear Regulatory Commission initiated research programs to study hydrogen behavior and control during accidents at nuclear plants. Several fundamental questions and issues arise when the hydrogen problem for light-water-reactor plants is examined. These relate to four aspects of the problem: hydrogen production; hydrogen transport, release, and mixing; hydrogen combustion; and prevention or mitigation of hydrogen combustion. Although much has been accomplished, some unknowns and uncertainties still remain, for example, the rate of hydrogen production during a degraded-core or molten-core accident, the rate of hydrogen mixing, the effect of geometrical structures and scale on combustion, flame speeds, combustion completeness, and mitigation-scheme effectiveness. This article discusses the nature and extent of the hydrogen problem, the progress that has been made, and the important unresolved questions

  1. Feedwater processing method in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izumitani, M; Tanno, K

    1976-09-06

    The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

  2. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  3. Industry water use : innovations, changes and challenges

    International Nuclear Information System (INIS)

    Braun, B.

    2004-01-01

    This paper presents work conducted by Canadian Natural Resources Ltd. (CNRL) in developing the McMurray Formation as an alternate water source in the Cold Lake Beaver River (CLBR) basin. Industry relies on both fresh water and brackish water to produce oil from thermal oil sands projects. A long-term sustainable supply of water is critical to the development of such projects. Although historically water has been considered as a renewable resource, it is currently viewed in a wider context. Technical advancements have made it possible to use recycled water for thermal recovery. Many heavy thermal oil expansions use brackish water. Capital costs are higher but heating costs are lower because brackish water is already warm. The use of brackish water allows companies to survive within their licenses while increasing production. Other possibilities include the use of depleted reservoir sections to store water to increase the use of recycled water. It was noted that brackish water resources need to be mapped and understood in greater detail. The objective is to use brackish water at a cost equal to, or less than fresh water. tabs., figs

  4. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

    The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

  5. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Juhn, P.E.

    1999-01-01

    Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

  6. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  7. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  8. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  9. Reprocessing technology for present water reactor fuels

    International Nuclear Information System (INIS)

    McMurray, P.R.

    1977-01-01

    The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

  10. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  11. Corrosion problems in light water nuclear reactors

    International Nuclear Information System (INIS)

    Berry, W.E.

    1984-01-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

  12. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  13. A review of the UKAEA interest in heavy water reactors

    International Nuclear Information System (INIS)

    Symes, R.J.

    1983-01-01

    The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

  14. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  15. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

    1994-01-01

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  16. Implications of Frugal Innovations on Sustainable Development: Evaluating Water and Energy Innovations

    Directory of Open Access Journals (Sweden)

    Jarkko Levänen

    2015-12-01

    Full Text Available Frugal innovations are often associated with sustainable development. These connections, however, are based on anecdotal assumptions rather than empirical evidence. This article evaluates the sustainability of four frugal innovations from water and energy sectors. For the purposes of the evaluation, a set of indicators was developed. Indicators are drawn from sustainable development goals by the United Nations and they encompass central dimensions of sustainability: ecological, social and economic. In this article, frugal innovations are compared to solutions that are currently used in similar low-income contexts. Studied frugal innovations were found more sustainable in terms of energy production and water purification capacity than the existing solutions. In terms of social sustainability, larger differences between innovations were found. For example, business models of frugal energy solutions focus on capacity building and the inclusion of marginalized low-income people, whereas business models of water purification solutions focus on more traditional corporate social responsibility activities, such as marketing awareness campaigns and cooperation with non-governmental organizations. Three major sustainability challenges for frugal innovators were identified: (1 the proper integration of material efficiency into product or service systems; (2 the patient promotion of inclusive employment; and (3 the promotion of inclusive and sustainable local industrialization. The article concludes that despite indisputable similarities between frugality and sustainability, it is problematic to equate the two conceptually.

  17. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Development of light water reactors and subjects for hereafter

    International Nuclear Information System (INIS)

    Murao, Yoshio

    1995-01-01

    As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

  20. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.) [de

  1. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  2. Innovative microbial fuel cell for electricity production from anaerobic reactors

    DEFF Research Database (Denmark)

    Min, Booki; Angelidaki, Irini

    2008-01-01

    A submersible microbial fuel cell (SMFC) was developed by immersing an anode electrode and a cathode chamber in an anaerobic reactor. Domestic wastewater was used as the medium and the inoculum in the experiments. The SMFC could successfully generate a stable voltage of 0.428 ± 0.003 V with a fixed......, a large portion of voltage drop was caused by the ohmic (electrolyte) resistance of the medium present between two electrodes, although the two electrodes were closely positioned (about 3 cm distance; internal resistance = 35 ± 2 Ω). The open circuit potential (0.393 V vs. a standard hydrogen electrode...

  3. Suppression device for the reactor water level lowering

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To suppress the lowering in the reactor water level so as to avoid unnecessary actuation of ECCS upon generation of transient changes which forecasts the lowering of the reactor water level in a BWR type reactor. Constitution: There are provided a water level suppression signal generator for generating a water level suppression signal upon generation of a transient change signal which forecasts the water level lowering in a nuclear reactor and a recycling flow rate controller that applies a recycling flow rate control signal to a recycling pump drive motor by the water level lowering suppression signal. The velocity of the recycling pump is controlled by a reactor scram signal by way of the water level lowering suppresion signal generator and a recycling flow rate controller. Then, the recycling reactor core flow rate is decreased and the void amount in the reactor is transiently increased where the water level tends to increase. Accordingly, the water level lowering by the scram is moderated by the increasing tendency of the water level. (Ikeda, J.)

  4. Innovations in Equipment Erection of Prototype Fast Breeder Reactor (PFBR)

    International Nuclear Information System (INIS)

    Sreekanth, S.; Kumar, Prabhat

    2013-01-01

    • PFBR equipment erection was a challenging task where thin walled vessels had transported and handled with utmost precautions to avoid redial forces on the vessels, which could buckle the vessels. • There was a real challenge in lifting the vessels without swing, placement of large size and heavy vessel at a distance of 57 meters where the crane operator has no line of sight to equipment's being erected. • Lot of care had been taken during lifting, handling and erection of thin walled ODC with innovative methods used for lifting fixtures, guiding arrangements, alignment fixtures and achieved the stringent erection tolerances

  5. Fuel density effect on parameter of reactivity coefficient of the Innovative Research Reactor core

    International Nuclear Information System (INIS)

    Rokhmadi; Tukiran S

    2013-01-01

    The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR) and then as an exchanger for old research reactor. The aim of this research is to complete RRI core design data as a requirement for design license. Calculation done is to get the RRI core reactivity coefficients with 5 x 5 core configuration and 20 MW of power, has more than 40 days cycle of length. The RRI core reactivity coefficient calculation is done for new U-"9Mo-Al fuel with variation of densities. The calculation is done by using WIMSD-5B and BATAN-FUEL computer codes. The result of calculation for conceptual design showed that the equilibrium RRI core with 5 x 5 configuration, 450 g, 550 g and 700 g of fuel loadings have negative reactivity coefficients of fuel temperature, moderator temperature, void fraction and density of moderator but the values of the reactivities are very variation. This results has met the safety criteria for RRI core conceptual design. (author)

  6. Research Reactors for the Development of Materials and Fuels for Innovative Nuclear Energy Systems

    International Nuclear Information System (INIS)

    2017-01-01

    This publication presents an overview of research reactor capabilities and capacities in the development of fuels and materials for innovative nuclear reactors, such as GenIV reactors. The compendium provides comprehensive information on the potential for materials and fuel testing research of 30 research reactors, both operational and in development. This information includes their power levels, mode of operation, current status, availability and historical overview of their utilization. A summary of these capabilities and capacities is presented in the overview tables of section 6. Papers providing a technical description of the research reactors, including their specific features for utilization are collected as profiles on a CD-ROM and represent an integral part of this publication. The publication is intended to foster wider access to information on existing research reactors with capacity for advanced material testing research and thus ensure their increased utilization in this particular domain. It is expected that it can also serve as a supporting tool for the establishment of regional and international networking through research reactor coalitions and IAEA designated international centres based on research reactors.

  7. Defect and Innovation of Water Rights System

    Institute of Scientific and Technical Information of China (English)

    Zhou Bin

    2008-01-01

    The rare deposition of water resources conflicts with its limitless demand. This determined the existence of the water rights transaction system. The implementation of the water rights transaction system requires clarifying the definition of water re-source fight above all distinctly. At present, it is a kind of common right system arrangement which needs the Chinese government to dispose of water resources. Though a series of management sys-tems guaranteed the government's supply of water resource, it hindered the development of the water market seriously and caused the utilization of water resources to stay in the inefficient or low efficient state for a long time. Thus, we should change the government's leading role in the resource distribution and really rely on the market to carry on the water rights trade and transac-tion. In this way, the water rights could become a kind of private property right relatively, and circulate freely in the market. As a result of this, we should overcome the defects of common right, make its external performance internalized maximally and achieve the optimized water resource disposition and use it more effec-tively.

  8. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  9. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  10. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  11. ORNL TNS Program: plasma engineering considerations and innovations for a medium field tokamak fusion reactor

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Attenberger, S.E.; Houlberg, W.A.; Mense, A.T.; Rome, J.A.; Uckan, N.A.

    1977-12-01

    Recent plasma engineering studies have ascertained a viable concept for The Next Step (TNS) reactor based on medium toroidal fields between 4 T and 7 T at the plasma center, plasma anti β values up to 10%, and averaged densities between 0.6 x 10 14 cm -3 and 2.5 x 10 14 cm -3 . Plasma engineering innovations that can substantially reduce the size, cost, and complexity of the TNS reactor have been explored and are summarized. It is shown that the previously anticipated requirement of high pellet velocities can be substantially reduced; the toroidal field (TF) ripple requirements may be relaxed to reduce the number of TF coils and improve machine access; hybrid equilibrium field (EF) coils have been shown to require building only small interior coils and to reduce the power supply required by the exterior coils; proper approaches of microwave plasma preheating may reduce the peak loop voltage for start-up by an order of magnitude. The medium-field TNS reactor concepts and the plasma engineering innovations discussed should be applicable to other designs of tokamak reactors; some of the suggested innovations will be tested in upcoming experiments

  12. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  13. An improved water cooled nuclear reactor and pressuriser assembly

    International Nuclear Information System (INIS)

    Gardner, F.J.; Strong, R.

    1991-01-01

    A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

  14. Research and development of super light water reactors and super fast reactors in Japan

    International Nuclear Information System (INIS)

    Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

    2011-01-01

    Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

  15. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

    Matal, O.; Sousek, P.; Simo, T.; Lehota, M.; Lipka, J.; Slugen, V.

    2000-01-01

    Defects in feed water distributing system due to corrosion-erosion effects have been observed at many VVER 440 steam generators (SG). Therefore analysis of defects origin and consequently design development and testing of a new feed water distributing system were performed. System tests in-situ supported by calculations and comparison of measured and calculated data were focused on demonstration of long term reliable operation, definition of water flow and water chemical characteristics at the SG secondary side and their measurements and study of dynamic characteristics needed for the innovated feed water distributing system seismic features approval. The innovated feed water distributing system was installed in the SGs of two VVER units already. (author)

  16. Summary report of the 7th reduced-moderation water reactor workshop

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

    2005-08-01

    As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

  17. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  18. Auxiliary water supply device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

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

  19. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    Brand, P.; Wiechers, W.K.

    1974-01-01

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  20. Transmutation of Americium in Light and Heavy Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

    2009-06-15

    There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

  1. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  2. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  3. A case study for INPRO methodology based on Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Anantharaman, K.; Saha, D.; Sinha, R.K.

    2004-01-01

    Under Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a methodology (INPRO methodology) has been developed which can be used to evaluate a given energy system or a component of such a system on a national and/or global basis. The INPRO study can be used for assessing the potential of the innovative reactor in terms of economics, sustainability and environment, safety, waste management, proliferation resistance and cross cutting issues. India, a participant in INPRO program, is engaged in a case study applying INPRO methodology based on Advanced Heavy Water Reactor (AHWR). AHWR is a 300 MWe, boiling light water cooled, heavy water moderated and vertical pressure tube type reactor. Thorium utilization is very essential for Indian nuclear power program considering the indigenous resource availability. The AHWR is designed to produce most of its power from thorium, aided by a small input of plutonium-based fuel. The features of AHWR are described in the paper. The case study covers the fuel cycle, to be followed in the near future, for AHWR. The paper deals with initial observations of the case study with regard to fuel cycle issues. (authors)

  4. Alternative water injection device to reactor equipment facility

    International Nuclear Information System (INIS)

    Yamashita, Masahiro.

    1995-01-01

    The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

  5. Integral approach to innovative fuel and material investigations in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.

    2009-01-01

    Integral approach used for fuel and material investigations in the Halden reactor can be used in support of qualification and certification of fuel to be introduced in commercial NPPs. This approach has been partly used for WWER fuel investigation in the Halden Reactor in a series of irradiation tests. In-pile fuel performance tests with reliable measurements provided by Halden instrumentation under different conditions can be used for validation of the WWER fuel behaviour models and verification of fuel performance codes. These models and codes can be used for qualification of innovative fuel behaviour under extended conditions

  6. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    Energy Technology Data Exchange (ETDEWEB)

    M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

    2005-04-26

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.

  7. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    International Nuclear Information System (INIS)

    Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

    2005-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications

  8. A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

    International Nuclear Information System (INIS)

    Choi, H.; Downar, T.J.

    1999-01-01

    A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

  9. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  10. Innovative nuclear reactor development. Opportunities for international co-operation

    International Nuclear Information System (INIS)

    2002-08-01

    A number of countries wish to expand their use of nuclear energy or keep open the option of doing so in the future. Any new nuclear generating capacity will be built in the context of increasingly privatized and de-regulated energy markets coupled with heightened public concern over nuclear power. New nuclear power plants must maintain or exceed current levels of safety and must be economically competitive with alternative ways of generating electricity. They must address other challenges as well, among them waste disposal and nonproliferation concerns. This report reviews how some of the innovative nuclear-fission technologies being developed today attempt to address the challenges facing nuclear energy. It suggests some areas for collaborative research and development that could reduce the time and cost required to develop new technologies. The report is a product of the 'Three-Agency Study', a joint project among the International Energy Agency (IEA), the OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA). (authors)

  11. Grey water treatment in UASB reactor at ambient temperature.

    Science.gov (United States)

    Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

    2007-01-01

    In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

  12. Innovative Sustainable Water Management Practices in Solar Residential Design

    Directory of Open Access Journals (Sweden)

    C. Jason Mabry

    2012-11-01

    Full Text Available This paper communicates the results of an architectural research project which sought innovative design strategies for achieving energy and resource efficiencies in water management systems traditionally used in single-family housing. It describes the engineering of an efficient, multifaceted, and fully integrated water management system for a domesticenvironment of 800 sq. ft., entirely powered by solar energy. The four innovations whose details are conveyed include the use of alternate materials for piping distribution and collection, the use of water in solar energy generation, the design of a building skin which capitalizes on water’s capacity to store heat as well as the design of a ecological groundscape which re-usesand filters waste water and rain water.Keywords: energy, plumbing, home design

  13. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  14. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  15. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  16. Overview of EU research activities in transmutation and innovative reactor systems within the Euratom framework programmes

    International Nuclear Information System (INIS)

    Bhatnagar, V.

    2009-01-01

    European Community (EC) (currently 27 Member States) shared-cost research has been organised in Framework Programmes (FP) of durations of 4 - 5 years since 1984. The 6th European Atomic Energy Community (EURATOM) Framework Programme (2002 - 06) and the current 7th FP (2007 - 11) have been allocated a fission research budget respectively of 209 and 287 Million Euro from the EC. There are 10 projects (total budget 70 M Euro, EC contribution 38 M Euro) in all aspects of transmutation ranging from road-mapping exercise to large integrated projects on accelerator driven systems, lead-cooled fast critical systems for waste transmutation, technology, fuel, accelerator facilities for nuclear data etc. In Innovative Reactor concepts, there are about half-a-dozen projects (total budget 30 M Euro, EC contribution 16 M Euro) including High Temperature Reactors, Gas-cooled Fast reactors, road-mapping exercise on sodium fast reactors etc. The main research and training activities in FP7 are: management of radioactive waste, reactor systems, radiation protection, infrastructures, human resources and mobility and training. In the two call for proposals (2007 and 2008) in FP7, 8 projects have been accepted in transmutation and innovative reactor concepts (total budget 53 M Euro, EC contribution 32 M Euro). These research projects cover activities ranging from materials, fuels, treatment of irradiated graphite waste, European sodium fast reactor to the establishment of a Central Design Team of a fast-spectrum transmutation device in Europe. The third call for proposals is underway requesting proposals on nuclear data, thermal hydraulics, gas and lead-cooled fast reactor systems with a total EC budget of 20 M Euro. International collaboration is an important element of the EU research policy. This overview paper will present elements of the strategy of EURATOM research and training in waste management including accelerator driven transmutation systems and Innovative reactor concepts

  17. Auxiliary equipment for cooling water in a reactor

    International Nuclear Information System (INIS)

    Konno, Yasuhiro; Sakairi, Toshiaki.

    1975-01-01

    Object: To effectively make use of pressure energy of reactor water, which has heretofore been discarded, to enable supply of emergency power supply of high reliability and to prevent spreading of environmental contamination. Structure: Sea water pumped by a sea water supply pump is fed to a heat exchanger. Reactor water carried through piping on the side to be cooled is removed in heat by the heat exchanger to be cooled and returned, and then again returned to the reactor. On the other hand, sea water heated by the heat exchanger is fed to a water wheel to drive the water wheel, after which it is discharged into a discharging path. A generator may be directly connected to the water wheel to use the electricity generated by the generator as the emergency power source. (Kamimura, M.)

  18. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  19. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  20. Material and water chemistry for a ferritic reactor coolant system in pressure water reactors

    International Nuclear Information System (INIS)

    Stieding, L.

    1979-04-01

    The use of unplated, low-alloy steels in a boric acid controlled PWR is not considered possible without changing the water conditions during the start-up and shut-down periods of the reactor. The significant pH reduction of the water due to boric acid during these periods most probably leads to damage of the magnetite protective layers followed by selective corrosion. As this highly important process has not been sufficiently evaluated with respect to our specific application problem, more detailed information will be necessary. KWU test facilities provide a means of performing such tests. In order to avoid corrosion attack during the above operating conditions, an inhibition of the water with 7 Li-borate is recommended which, however, will amount to approx. DM 60.000,-- per period of use. (orig.) [de

  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. RETRAN sensitivity studies of light water reactor transients. Final report

    International Nuclear Information System (INIS)

    Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

    1977-06-01

    This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

  3. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  4. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    Mattern, J.

    1976-01-01

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

  5. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-01-01

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  6. Calculations for accidents in water reactors during operation at power

    International Nuclear Information System (INIS)

    Blanc, H.; Dutraive, P.; Fabrega, S.; Millot, J.P.

    1976-07-01

    The behaviour of a water reactor on an accident occurring as the reactor is normally operated at power may be calculated through the computer code detailed in this article. Reactivity accidents, loss of coolant ones and power over-running ones are reviewed. (author)

  7. Overview of environmental materials degradation in light-water reactors

    International Nuclear Information System (INIS)

    Shaaban, H.I.; Wu, P.

    1986-08-01

    This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

  8. 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

  9. Small high temperature gas-cooled reactors with innovative nuclear burning

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Ismail; Sekimoto, Hiroshi

    2008-01-01

    Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles. (author)

  10. 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

  11. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  12. A gap analysis of the South African innovation system for water ...

    African Journals Online (AJOL)

    A gap analysis of the South African innovation system for water. ... Two major approaches to science and innovation from the innovation systems ... infrastructure and data sharing; reorganising the research environment within universities; ...

  13. Guidance for the evaluation of innovative nuclear reactors and fuel cycles. Report of Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2003-06-01

    The IAEA General Conference in 2000 invited all interested Member States to combine their efforts under the aegis of the IAEA in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology. Resolutions of the UN General Assembly in 2001 and 2002 provided additional endorsement for INPRO, by emphasizing the unique role that the IAEA can play in developing user requirements and in addressing safeguards, safety, and environmental questions for innovative reactors and their fuel cycles and stressing the need for international collaboration in the development of innovative nuclear technology. As of April 2003, INPRO had 15 members: Argentina, Brazil, Bulgaria, Canada, China, Germany, India, Republic of Korea, Pakistan, Russian Federation, Spain, Switzerland, the Netherlands, Turkey and the European Commission. The main objectives of INPRO are to: Help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner; and to Bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles. The 21st century promises the most competitive, globalized markets in human history, the most rapid pace of technological change ever, and the greatest expansion of energy use, particularly in developing countries. For a technology to make a truly substantial contribution to energy supplies, innovation is essential. It will be the defining feature of a successful nuclear industry and a critical feature of international co-operation in support of that industry, co-operation that ranges from joint scientific and technological initiatives, to safety standards and guidelines, and to security and safeguards activities. Innovation is also essential to attract a growing, high-quality pool of talented scientists, engineers and

  14. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  15. Brief summary of water reactor fuel activities in China

    Energy Technology Data Exchange (ETDEWEB)

    Zhongyue, Zhang [China Inst. of Atomic Energy (CIAE), Beijing (China)

    1997-12-01

    The presentation briefly reviews the water reactor fuel activities in China describing: nuclear power development program and growth forecast; fuel performance;fuel performance code improvement; research and development plans. 1 ref., 3 figs, 2 tabs.

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

    International Nuclear Information System (INIS)

    Bouteille, F.; Schneider, D.

    2002-01-01

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

  17. Experience in operation of heavy water reactors

    International Nuclear Information System (INIS)

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

    The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

  18. The international project on innovative nuclear reactors and fuel cycles (INPRO) - status and trends

    International Nuclear Information System (INIS)

    Gowin, Peter J.; Beatty, Randy L.

    2010-01-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in 2000. As of April 2010, INPRO has 31 members and is implementing activities in five programme areas: A: Nuclear Energy System Assessments (NESA) using the INPRO Methodology Assisting Member States in performing Nuclear Energy System Assessments (NESA) using the INPRO methodology, in support of long-term strategic planning and nuclear energy deployment decision making. B: Global Vision Developing global and regional nuclear energy scenarios, on the basis of a scientific-technical pathway analysis, that lead to a global vision on sustainable nuclear energy development in the 21. century, and supporting Member States in working towards that vision. C: Innovations in Nuclear Technology Fostering collaboration among INPRO Member States on selected innovative nuclear technologies and related R and D that contribute to sustainable nuclear energy. D: Innovations in Institutional Arrangements Investigating and fostering collaboration on innovative institutional and legal arrangements for the use of innovative nuclear systems in the 21. century and supporting Member States in developing and implementing such innovative arrangements. E: INPRO Dialogue Forum Bringing together technology holders and technology users to discuss, debate and share information on desirable innovations, both technical and institutional, but also national long-term nuclear planning strategies and approaches and, on the highest level, the global nuclear energy system. The paper presents main INPRO achievements to date, the current status of activities in these five programme areas and recent INPRO publications, in particular in support of nuclear energy system assessments (NESA) using the INPRO methodology. (authors)

  19. Water and Regolith Shielding for Surface Reactor Missions

    Science.gov (United States)

    Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  20. Water and Regolith Shielding for Surface Reactor Missions

    International Nuclear Information System (INIS)

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

  1. Safety considerations concerning light water reactors in Sweden

    International Nuclear Information System (INIS)

    Nilsson, T.

    1977-01-01

    In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

  2. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  3. Analysis of an accelerator-driven subcritical light water reactor

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

    1997-01-01

    An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

  4. The Steam Generating Heavy Water Reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

  5. Pressurised water reactor fuel management using PANTHER

    International Nuclear Information System (INIS)

    Parks, G.T.; Knight, M.P.

    1996-01-01

    This paper describes the integration of Nuclear Electric's reactor physics code PANTHER with an automatic optimisation procedure designed to search for optimal PWR reload cores and assesses its performance. (Author)

  6. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  7. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  8. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  9. Innovation, adaptability, and collaboration: Keys to success for small and medium sized reactors. Cairo, 27 May 2001

    International Nuclear Information System (INIS)

    ElBaradei, M.

    2001-01-01

    Small and medium sized reactors, within a power output of less than 700 MW(e), are receiving increased consideration in the effort to meet changing market requirements. Smaller plants allow a more incremental investment, which can be used to hedge against demand uncertainty. They are more suitable for standardization and prefabrication, which in turn encourages enhanced quality control and stimulates rapid development of expertise and shorter construction schedules. They provide a better match to grid capacity in developing countries. And they are more easily adapted to a broad range of industrial settings and applications, such as district heating, heavy oil recovery, or the production of hydrogen and other chemical fuels. Sea water desalination is an application for which smaller reactors hold a particular advantage. Nuclear powered desalination is a proven technology. Clearly, we live in an era in which our society faces many difficult economic, environmental and social issues associated with sustainable development and energy demand. Against that backdrop, nuclear power is a mature technology that deserves careful consideration as a contributor to solving some of these issues. The development of innovative small and medium sized reactors will play a key role in helping to match state-of-the-art technology to user needs. An exchange of information and ideas is a step towards further progress

  10. Device for controlling water supply to nuclear reactor

    International Nuclear Information System (INIS)

    Iwasaki, Toshio.

    1974-01-01

    Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

  11. An innovative design approach to a cost effective commercial liquid metal reactor

    International Nuclear Information System (INIS)

    Carelli, M. D.; Garkisch, H. D.; Hundal, R.; Arie, K.; Handa, N.; Ota, H.; Matsuyama, S.; Todreas, N.; Hejzlar, P.; Wells, P.; Louis, T. S.

    2008-01-01

    One of the requirements for the Advanced Recycle Reactor (ARR) in the DOE GNEP program is to be economically competitive with advanced LWRs. Sodium cooled Liquid Metal Reactors have been designed in two basic configurations, pool and loop, and various designs had different features, but the capital cost has consistently been substantially higher than for LWRs in spite of the LMRs higher efficiency due to the higher coolant exit temperature. We have identified two major areas as responsible for the divergence in capital costs. First, LMRs have been designed within the HCDA (Hypothetical Core Disruptive Accident) philosophy. Basically, the plant has to survive a major accident regardless whether this accident has any mechanistically acceptable, albeit very small, probability of occurring. Thus, in spite of their inherent reactivity feedbacks, much more effective than in LWRs, LMRs have been designed with redundant systems and costly design solutions which a probabilistic, rather than a deterministic approach, would eliminate or greatly simplify. The other major consideration is the presence of the intermediate heat transport system to prevent a Na/H 2 O reaction involving primary (radioactive) sodium. No other reactor type has three heat transport systems and in fact the LMR intermediate heat transport system is a substantial contributor to the capital cost. For our ARR rated at 1000 MWt power, we have selected a pool configuration because of its enhanced safety and its inherent capability of yielding a more compact reactor. The intermediate heat transport system is eliminated. Immersed in the pool is a double wall steam generator where water/steam flows through the inner tubes and Helium is in the space between the inner and outer tubes for detection purposes. Models of two different types of double wall steam generators designed independently by Westinghouse and Toshiba had been tested in the 1970-1980's at ETEC with favorable results. As we fully realize the risks

  12. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  13. Effect of water impurities on stress corrosion cracking in a boiling water reactor

    International Nuclear Information System (INIS)

    Ljungbery, L.G.; Cubicciotti, D

    1985-01-01

    A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

  14. Device for preventing cooling water from flowing out of reactor

    International Nuclear Information System (INIS)

    Chinen, Masanori; Kotani, Koichi; Murase, Michio.

    1976-01-01

    Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

  15. Analysis of thermal fatigue events in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

  16. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

    Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

    1997-01-01

    The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

  17. The international project on innovative nuclear reactors and fuel cycles (INPRO): status and outlook

    International Nuclear Information System (INIS)

    Steur, R.; Kupitz, J.; Depisch, F.

    2004-01-01

    Full text: During the last fifty years remarkable results are achieved in the application of nuclear technology for the production of electricity. Looking ahead to the next fifty years it is clear that the demand for energy will grow considerably and also new requirements for the way the energy will be supplied have to be fulfilled. Following a resolution of the General Conference of the IAEA in the year 2000 an International Project on Innovative Nuclear Reactors and Fuel Cycles, referred to as INPRO, was initiated. The main objectives of INPRO are to: Help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner; and Bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles. Within INPRO the future of the energy demand and supply was explored and several scenario's identified. A leading requirement for energy supply is coming up and will play a crucial role: sustainability of the way the energy supply will be realized. Fulfilling the growing need for energy in developing countries is as well an important issue. Based on these scenario's for the next fifty years, requirements for the different aspects of the future of nuclear energy systems, such as economics, sustain ability and environment, safety, waste and proliferation resistance have been identified as well a methodology developed. to assess innovative nuclear systems and fuel cycles. On the base of this assessment, the need for innovations and breakthroughs in existing technology can be defined. To facilitate the deployment of innovative nuclear systems also different aspects of the infrastructure, technical as well institutional have been reviewed and recommendations for changes are made to anticipate main developments in the world such as the ongoing globalisation. As a contribution to the conference

  18. Automatic radiometric analyzer for nuclides in nuclear reactor water

    International Nuclear Information System (INIS)

    Kitamura, Masao; Tokoi, Hiromi; Kitaguchi, Hiroshi; Ozawa, Yoshihiro; Urata, Megumu.

    1981-01-01

    Purpose: To shorten the processing time and improve the accuracy for processing water sampled from reactor coolants, as well as simplify the mechanism of the apparatus. Constitution: Reactor water sampled from reactor coolants, after filtered out with insoluble solids, is stored in an ion exchange container. Thereafter, the amount of ion exchanged water is regulated by the coarse measurement of radioactivity concentration by a monitor. Further, ion exchange resins are charged from a resin tank, agitated by gases and dispersed into sampled water. Then, all of the radioactive iodines contained in the sample are collected in the resins. The resins are recovered through evacuation into instrumenting vessels for measurement of radioactivity. Since ion exchange resins are dispersed in the sampled water in this system, the processing time can be shortened. (Ikeda, J.)

  19. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    PetrusTakaki, N.

    2012-01-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  20. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  1. Business model innovation in the water sector in developing countries.

    Science.gov (United States)

    Gebauer, Heiko; Saul, Caroline Jennings

    2014-08-01

    Various technologies have been deployed in household devices or micro-water treatment plants for mitigating fluoride and arsenic, and thereby provide safe and affordable drinking water in low-income countries. While the technologies have improved considerably, organizations still face challenges in making them financially sustainable. Financial sustainability questions the business models behind these water technologies. This article makes three contributions to business models in the context of fluoride and arsenic mitigation. Firstly, we describe four business models: A) low-value devices given away to people living in extreme poverty, B) high-value devices sold to low-income customers, C) communities as beneficiaries of micro-water treatment plants and D) entrepreneurs as franchisees for selling water services and highlight the emergence of hybrid business models. Secondly, we show current business model innovations such as cost transparency & cost reductions, secured & extended water payments, business diversification and distribution channels. Thirdly, we describe skills and competencies as part of capacity building for creating even more business model innovations. Together, these three contributions will create more awareness of the role of business models in scaling-up water treatment technologies. Copyright © 2014 Elsevier B.V. All rights reserved.

  2. International trend on development of an innovative nuclear reactor and its meanings

    Energy Technology Data Exchange (ETDEWEB)

    Matsui, Kazuaki [Institute of Applied Energy, Tokyo (Japan)

    2002-01-01

    On outlining on flow of so-called innovative or new type nuclear reactor, at first, an improvement line of large-scale WHR, contains ABWR-2, APWR and its successive APWR+ in Japan, APR in Korea, and EPR in Europe, all of which have super large-scale output of 1.5MKW to use their scale merits in maximum. And, the second type is fast reactor only in Russia and Japan which are under reviewing its actual using plan of its already established development route. Furthermore, nuclear industry in the world is allowable to say a has-been industry, even its R and D system is decrepit, its researchers are much aged, and even utilization and foreign development of nuclear energy as a protecting measure of global warming are pronounced its self-control at the Bonn Conference in last year. However, the Generation 4 International Forum led by U.S.A. since early of 2000 and the Innovative Reactor Development Program (INPRO) through the International Atomic Energy Association (IAEA) due to initiative of Russia are planned to cooperatively promote their programs. In order to obtain any priority on small-scale production considerable technical jump is required or R and D and technical development elements with technical gap is necessary, which must be proved establishment of a target to overcome their scale demerit. (G.K.)

  3. International Project on Innovative Nuclear Reactors and Fuel Cycles: Introduction and Education and Training Activity

    International Nuclear Information System (INIS)

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

    2015-01-01

    The IAEA’s International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was established in 2000 through IAEA General Conference resolution with aim to ensure that sustainable nuclear energy is available to help meet the energy needs of the 21st century. INPRO seeks to bring together technology holders, users and newcomers to consider jointly the international and national actions required for achieving desired innovations in nuclear reactors and fuel cycles, with a particular focus on sustainability and needs of developing countries. It is a mechanism for INPRO Members to collaborate on topics of joint interest. INPRO activities are undertaken in close cooperation with Member States in the following main areas: Global Scenarios, Innovations, Sustainability Assessment and Strategies, Policy and Dialogue. The paper presents short introduction in INPRO and specifically the distant Education and Training INPRO activity on important topics of nuclear energy sustainability to audiences in different Member States. These activities can support capacity building and national human resource development in the nuclear energy sector. The main benefit of such training courses and workshops is that it is not only targeted to students, but also to lecturers of technical and nuclear universities. Moreover, young professionals working at nuclear energy departments, electric utilities, energy ministries and R&D institutions can participate in such training and benefit from it. (authors)

  4. The IAEA international project on innovative nuclear reactors and fuel cycles (INPRO): current and future activities

    International Nuclear Information System (INIS)

    Kupitz, J.; Depisch, F.; Kuznetsov, V.

    2004-01-01

    Upon resolutions of the IAEA General Conference in 2000, the IAEA initiated International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). The objective of INPRO, which comprises two phases, is to support sustainable deployment and use of nuclear technology to meet the global energy needs in the next 50 years and beyond. During Phase I, work is subdivided into two sub phases. Phase 1A focused on determining user requirements in the areas of economics, environment, safety, proliferation resistance, and recommendations in the area of so-called crosscutting issues, which are legal, institutional, and infrastructure issues accompanying the deployment of nuclear power, and is targeted at developing a methodology and guidelines for the assessment of various nuclear reactor and fuel cycle concepts and approaches. Phase 1A was finalised in June 2003 with its results now available as IAEA TECDOC-1362. Phase 1B has started in July 2003. During this phase interested Member States are performing case studies to validate the INPRO methodology and, later on, to assess selected innovative nuclear energy systems using the updated INPRO methodology. In accordance with the INPRO Terms of Reference, after successful completion of Phase I, Phase II may be initiated to examine the feasibility of commencing international projects on innovative nuclear energy systems. The paper contains a description of the current and future activities of INPRO and summarizes the outcome of the project.(author)

  5. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  6. Assessment of two small-sized innovative nuclear reactors for electricity generation in Brazil using INPRO methodology

    International Nuclear Information System (INIS)

    Goncalves Filho, Orlando Joao Agostinho; Sefidvash, Farhang

    2009-01-01

    This paper presents the main results of the assessment study of two small-sized innovative reactors for electricity generation in Brazil using the methodology developed under the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), co-ordinated by the International Atomic Energy Agency (IAEA). INPRO was initiated in 2001 and has the main objective of helping to ensure that nuclear energy is available to contribute in a sustainable manner to the energy needs of the 21st century. Brazil joined the INPRO project since its beginning and in 2005 submitted a proposal for the assessment using INPRO methodology of two small-sized reactors (IRIS - International Reactor Innovative and Secure, and FBNR - Fixed Bed Nuclear Reactor) as potential components of an innovative nuclear energy system (INS) completed by a conventional open nuclear fuel cycle based on enriched uranium. The scope of this assessment study was restricted to the reactor component of the INS and to the methodology areas of economics and safety for IRIS, and proliferation resistance and safety for FBNR. The results indicate that both IRIS and FBNR innovative designs comply mostly with the basic principles of the areas assessed and have potential to comply with the remaining ones. (author)

  7. The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    Juergen Kupitz

    2002-01-01

    This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited 'all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology' (GC(44)/RES/21) and invited Member States to consider to contribute to a task force on innovative nuclear reactors and fuel cycle (GC(44)/RES/22). In response to this invitation, the IAEA initiated an 'International Project on Innovative Nuclear Reactors and Fuel Cycles', INPRO. The Terms of Reference for INPRO were adopted at a preparatory meeting in November 2000, and the project was finally launched by the INPRO Steering Committee in May 2001. At the General Conference in 2001, first progress was reported, and the General Conference adopted a resolution on 'Agency Activities in the Development of Innovative Nuclear Technology' [GC(45)/RES/12, Tab F], giving INPRO a broad basis of support. The resolution recognized the 'unique role that the Agency can play in international collaboration in the nuclear field'. It invited both 'interested Member States to contribute to innovative nuclear technology activities' at the Agency as well as the Agency itself 'to continue it's efforts in these areas'. Additional endorsement came in a UN General Assembly resolution in December 2001 (UN GA 2001, A/RES/56/94), that again emphasized 'the unique role that the Agency can play in developing user requirements and in addressing safeguards, safety and environmental questions for innovative reactors and their fuel cycles' and stressed 'the need for international collaboration in the development of innovative nuclear technology'. As of February 2002, the following countries or entities have become members of INPRO: Argentina

  8. Water Innovations and Lessons Learned From Water Recycling in Space

    Science.gov (United States)

    Flynn, Michael

    2013-01-01

    This Presentation will cover technology and knowledge transfers from space exploration to earth and the tourism industry, for example, water and air preservation, green buildings and sustainable cities.

  9. Pressurized heavy-water reactor safety

    International Nuclear Information System (INIS)

    Pease, L.; Wilson, R.

    1977-09-01

    CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

  10. Method of controlling power of a heavy water reactor

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1975-01-01

    Object: To adjust a level of heavy water in a region of reflection body to control power in a heavy water reactor. Structure: The interior of a core tank filled with heavy water is divided by a partition into a core heavy water region and a reflection body region formed by surrounding the core heavy water region, and a level of heavy water within the reflection body region is adjusted to control power. Preferably, it is desirable to communicate the core heavy water region with the reflection body heavy water region at their lower portion, and gas pressure applied to an upper portion within at least one of said regions is adjusted to adjust the level of heavy water within the reflection body heavy water region. Thereby, the heavy water within the reflection body heavy water region may be introduced into the core region, thus requiring no tank which stores heavy water within the reflection body region. (Kamimura, M.)

  11. Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

    International Nuclear Information System (INIS)

    Hecker, H.C.; Freeman, L.B.

    1981-08-01

    This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

  12. 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

  13. Flow analysis in a supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Oh, C.H.; Kochan, R.J.; Beller, J.M.

    1996-01-01

    Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

  14. Safety of light water reactors. Risks of nuclear technology

    International Nuclear Information System (INIS)

    Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

    2012-01-01

    The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

  15. Development status and application prospect of supercritical-pressure light water cooled reactor

    International Nuclear Information System (INIS)

    Li Manchang; Wang Mingli

    2006-01-01

    The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

  16. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

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

    International Nuclear Information System (INIS)

    2004-10-01

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

  18. Biological mine water treatment operating a one stage reactor system

    CSIR Research Space (South Africa)

    Baloyi, MJ

    2006-05-01

    Full Text Available rumen fluid as source of the fermentation organisms, were utilised as electron donor when sulphate, as the electron acceptor, is converted to sulphide. The feed water entered the reactor at the top, to allow the water to get in contact with grass...

  19. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics and Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Anderson, Mark; Corradini, M.L.; Sridharan, K.; Wilson, P.; Cho, D.; Kim, T.K.; Lomperski, S.

    2004-01-01

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers

  20. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

    1994-08-01

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  1. The French Fast Reactor Program - Innovations in Support to Higher Standards

    International Nuclear Information System (INIS)

    Gauché, François

    2013-01-01

    • From the experience of ASTRID first phase of conceptual design studies (2010-2012), two remarks can be made: → Higher requirements in safety and operability lead to higher costs that cannot be fully recovered by advances in technology. This puts additional pressure on the next phases of the design to optimize the design and to keep the costs to the minimum. → There is a clear link between the level of safety that can be achieved and the maturity of the technology, i.e. the experience accumulated in R&D, design, construction, operation and decommissioning of past reactors. In the field of fast neutron reactors, this gives a strong advantage to the sodium technology, because strengths and weaknesses are well mastered. • Meeting the high requirements set for ASTRID and serving R&D needs of innovative options will require increased industrial and international collaboration

  2. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  3. International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). 2008 progress report

    International Nuclear Information System (INIS)

    2009-02-01

    The purpose of the work is to review the progress of the IAEA international project for innovative reactors and fuel cycle technologies (INPRO). The publication reports about the recognition of INPRO and on general Information on INPRO, its strengths, memberships, collaboration with other international initiatives, the INPRO organization and management and the history of INPRO. The section on the progress of INPRO in 2008 contains task 1: INPRO Methodology, task 2: Assessment Studies, task 3: Nuclear Energy Visions for the 21st Century, task 4: Infrastructure and Institutional Innovation, task 5: Common User Considerations and task 6: Collaborative Projects. Conclusions and New Trends are followed by a bibliography. Annex I deals with the INPRO project management in 2008 and Annex II provides a selection of photographs from 2008. Finally a list of acronyms is provided

  4. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

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

    1985-11-01

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

  5. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

    Sponton, L.; Sjoeberg, A.; Nordlinder, S.

    2001-01-01

    This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

  6. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  7. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  8. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  9. Preliminary Design of Compressor Impeller for innovative Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jekyoung; Cho, Seongkuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For nuclear power plant application, applying S-CO{sub 2} Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO{sub 2} Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO{sub 2} Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO{sub 2} near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAIST{sub T}MD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately.

  10. Development of an innovative reflector drive mechanism using magnetic repulsion force for 4S reactor

    International Nuclear Information System (INIS)

    Tsuji, K.; Watanabe, M.; Inagaki, H.; Nishikawa, A.; Takahashi, H.; Wakamatsu, M.; Matsumiya, H.; Nishiguchi, Y.

    2001-01-01

    A small sized fast reactor 4S: (Super Safe Small and Simple) which has a core of 10 - 30 years life time is controlled by reflectors. The reflector is required to be risen at very low speed to make up for the reactivity swing during operation. This report shows the development of an innovative reflector drive mechanism using magnetic repulsion force that can move at a several micrometer per one step. This drive mechanism has a passive shut down capability, and can eliminate reflector drive line. (author)

  11. Innovated re-claimed communal water - Benefits and challenges

    Directory of Open Access Journals (Sweden)

    Kostandin Kristo

    2018-03-01

    Full Text Available The drought regarding a climate forecast in Albanian departments that have marked the past three months were the result of winter rainfall that was insufficient for replenishing water tables this year and before. It is often said that water shortages are not a threat for Albania due to plenty water sources. It’s possible to think of this compared to other countries. Still in some Albanian regions the availability of water resources is becoming an urgent problem, even more so if we link this to the different scenarios tied to climate change. Each basin has specificities, but some general evolutions should be planned for the increase in water demand due to the rise of temperatures causing an increase in the price of water. This paper aims to introduce the benefits and challenges of applying at homes or communal buildings an Eco-innovation through developing a new inside distribution of wastewater so that it concludes into sustainable development of communal water distribution, therefore recycling in houses. Another objective of the study is to make evident the cost - efficiency importance of these re-distribution systems and how they affect improvement in water needs sector, highlighting the deficiencies that cause their not fully-efficient re-use of grey inside water and the positive impact on potable water saving.

  12. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  13. Structural integrity of water reactor pressure boundary components

    International Nuclear Information System (INIS)

    Loss, F.J.

    1977-01-01

    The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

  14. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  15. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-15

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

  17. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  18. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  19. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  20. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  1. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  2. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  3. Method of operating a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Lysell, G.

    1975-01-01

    When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

  4. UK methods for studying fuel management in water moderated reactors

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1970-10-01

    Current UK methods for studying fuel management and for predicting the reactor physics performance for both light and heavy water moderated power reactors are reviewed. Brief descriptions are given of the less costly computer codes used for initial assessment studies, and also the more elaborate programs associated with detailed evaluation are discussed. Some of the considerations influencing the accuracy of predictions are included with examples of various types of experimental confirmation. (author)

  5. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  6. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  7. A `big-mac` high converting water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ronen, Y; Dali, Y [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering

    1996-12-01

    Currently an effort is being made to get rid of plutonium. Therefore, at this time, a scientific study of a high converting reactor seems to be out of place. However , it is our opinion that the future of nuclear energy lies, among other things in the clever utilization of plutonium. It is also our opinion that one of the best ways to utilize plutonium is in high converting water reactors (authors).

  8. Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor

    International Nuclear Information System (INIS)

    Moussiere, S.

    2006-12-01

    Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

  9. Design requirements for innovative homogeneous reactor, lesson learned from Fukushima accident

    Science.gov (United States)

    Arbie, Bakri; Pinem, Suryan; Sembiring, Tagor; Subki, Iyos

    2012-06-01

    The Fukushima disaster is the largest nuclear accident since the 1986 Chernobyl disaster, but it is more complex as multiple reactors and spent fuel pools are involved. The severity of the nuclear accident is rated 7 in the International Nuclear Events Scale. Expert said that "Fukushima is the biggest industrial catastrophe in the history of mankind". According to Mitsuru Obe, in The Wall Street Journal, May 16th of 2011, TEPCO estimates the nuclear fuel was exposed to the air less than five hours after the earthquake struck. Fuel rods melted away rapidly as the temperatures inside the core reached 2800 C within six hours. In less than 16 hours, the reactor core melted and dropped to the bottom of the pressure vessel. The information should be evaluated in detail. In Germany several nuclear power plant were shutdown, Italy postponed it's nuclear power program and China reviewed their nuclear power program. Different news come from Britain, in October 11, 2011, the Safety Committee said all clear for nuclear power in Britain, because there are no risk of strong earthquake and tsunami in the region. Due to this severe fact, many nuclear scientists and engineer from all over the world are looking for a new approach, such as homogeneous reactor which was developed in Oak Ridge National Laboratory in 1960-ies, during Dr. Alvin Weinberg tenure as the Director of ORNL. The paper will describe the design requirement that will be used as the basis for innovative homogeneous reactor. Innovative Homogeneous Reactor is expected to reduce core melt by two decades (4), since the fuel is intermix homogeneously with coolant and secondly we eliminate the used fuel rod which need to be cooled for a long period of time. In order to be successful for its implementation of the innovative system, testing and validation, three phases of development will be introduced. The first phase is Low Level Goals is really the proof of concept;the Medium Level Goal is Technical Goalsand the High

  10. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  11. Entropy Generation Minimization for Reverse Water Gas Shift (RWGS Reactors

    Directory of Open Access Journals (Sweden)

    Lei Zhang

    2018-05-01

    Full Text Available Thermal design and optimization for reverse water gas shift (RWGS reactors is particularly important to fuel synthesis in naval or commercial scenarios. The RWGS reactor with irreversibilities of heat transfer, chemical reaction and viscous flow is studied based on finite time thermodynamics or entropy generation minimization theory in this paper. The total entropy generation rate (EGR in the RWGS reactor with different boundary conditions is minimized subject to specific feed compositions and chemical conversion using optimal control theory, and the optimal configurations obtained are compared with three reference reactors with linear, constant reservoir temperature and constant heat flux operations, which are commonly used in engineering. The results show that a drastic EGR reduction of up to 23% can be achieved by optimizing the reservoir temperature profile, the inlet temperature of feed gas and the reactor length simultaneously, compared to that of the reference reactor with the linear reservoir temperature. These optimization efforts are mainly achieved by reducing the irreversibility of heat transfer. Optimal paths have subsections of relatively constant thermal force, chemical force and local EGR. A conceptual optimal design of sandwich structure for the compact modular reactor is proposed, without elaborate control tools or excessive interstage equipment. The results can provide guidelines for designing industrial RWGS reactors in naval or commercial scenarios.

  12. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    Reuss, Paul

    1979-10-01

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235 U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235 U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters [fr

  13. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2013-10-15

    Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  14. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  15. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  16. A water inner circulation device for a reactor vessel

    International Nuclear Information System (INIS)

    Eriksson, O.

    1976-01-01

    A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

  17. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    Persault, M.

    1977-10-01

    Gamma spectroscopy analysis of spent fuels in power reactors; study of two typical cases: determination of the power distribution by the mean of the activity of a low periodic element (Lanthanum 140) and determination of the burnup absolute rate by examining the ratio of Cesium 134 and Cesium 137 activities. Measures were realized on fuel solutions and on fuel assemblies. Development of a power distribution map of the assemblies and comparison with the results of a three dimensional calculation of core evolution [fr

  18. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane; Ait-Djoudi, Fariza; Naceur, Wahib M.; Ghaffour, NorEddine

    2016-01-01

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria

  19. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  20. The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO): General description and implications for the research reactor infrastructure needed for R and D

    International Nuclear Information System (INIS)

    Sokolov, Yury A.

    2005-01-01

    The substantial growth in 21st century energy supplies needed to meet sustainable development goals has been emphasized by UNCSD, WSSD, IPCC and others. This will be driven by continuing population growth, economic development and aspiration to provide access to modern energy systems to the 1,6 billion people now without such access, the growth demand on limiting greenhouse gas emissions, and reducing the risk of climate change. A key factor to the future of nuclear power is the degree to which innovative nuclear technologies can be developed to meet challenges of economic competitiveness, safety, waste and proliferation concerns. There are two major international initiatives in the area of innovative nuclear technology: the IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycle (INPRO) and the Generation IV International Forum. With INPRO some scenarios of future energy needs were identified and the methodology for holistic assessment of the innovative nuclear energy systems (INS), which can be developed to meet these scenarios, was developed.. The current status of the INPRO project and details of the INPRO methodology will be reported. The research needs identified due to Agency's activities on innovative nuclear system development assume the use of research reactors. The areas crucial for the development of INS which critically dependent of the RR experiments and following requirements addressed to the RR will be discussed. These areas include the development of advanced fuel and core materials for proposed innovative power reactor concepts. (author)

  1. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  2. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  3. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  4. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  5. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  6. Features, present condition of development and future scope on the high temperature gas reactor as an innovative one

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2001-01-01

    The high temperature gas reactor has some features without previous reactors such as high temperature capable of taking-out, high specific safety, feasibility adaptable to versatile fuel cycle, and so on. Then, it is expected to be an innovative reactor to contribute to diversification of energy supply and expansion of energy application field. In Japan, under the HTTR (high temperature engineering test reactor) plan, construction of HTTR, which is the first high temperature gas reactor in Japan, was finished and its output upgrading test has been promoted. And, on the HTTR plan, together with promotion of full power operation, reactor performance tests, safety proof test, and so on, it is planned to carry out study on application of the high temperature heat such as hydrogen production and so on to aim to practise establishment and upgrading of technologies on high temperature gas reactor in Japan. Here were introduced features and present condition of development of the high temperature gas reactor as an innovative type reactor and described role and future scope in Japan. (G.K.)

  7. Study and modelling of an innovative coprecipitation reactor for radioactive liquid wastes decontamination

    International Nuclear Information System (INIS)

    Flouret, Julie

    2013-01-01

    In order to decontaminate radioactive liquid wastes of low and intermediate levels, the coprecipitation is the process industrially used. The aim of this PhD work is to optimize the continuous process of coprecipitation. To do so, an innovative reactor is designed and modelled: the continuous reactor/classifier. Two model systems are studied: the coprecipitation of strontium by barium sulphate and the sorption of cesium by PPFeNi. The simulated effluent contains sodium nitrate in order to consider the high ionic strength of radioactive liquid wastes. First, each model system is studied on its own, and then a simultaneous treatment is performed. The kinetic laws of nucleation and crystal growth of barium sulphate are determined and incorporated into the coprecipitation model. Kinetic studies and sorption isotherms of cesium by PPFeNi are also performed in order to acquire the necessary data for process modelling. The modelling realised enables accurate prediction of the residual strontium and cesium concentrations according to the process used: it is a valuable tool for the optimization of existing units, but also the design of future units. The continuous reactor/classifier presents many advantages compared to the classical continuous process: the decontamination efficiency of strontium and cesium is highly improved while the volume of sludge generated by the process is reduced. A better liquid/solid separation is observed in the reactor/classifier and the global installation is significantly more compact. Thus, the radioactive liquid wastes treatment processes can be intensified by the continuous reactor/classifier, which represents a very promising technology for future industrial application. (author) [fr

  8. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  9. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  10. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  11. Sea water desalination using nuclear reactors

    International Nuclear Information System (INIS)

    Nisan, S.

    2003-01-01

    The paper first underlines the water shortage problem today and in the years to come when, around the time horizon 2020, two-thirds of the total world population would be without access to potable water. Desalination of sea-water (and, to a limited extent, that of brackish water) is shown to be an attractive solution. In this context, sea-water desalination by nuclear energy appears to be not only technically feasible and safe but also economically very attractive and a sustainable solution. Thus, compared to conventional fossil energy based sources, desalination costs by nuclear options could be 30 to 60% lower. The nuclear options are therefore expected to satisfy the fundamental water needs and electricity demands of human beings without in any way producing large amounts of greenhouse gases which any desalination strategy, based on the employment of fossil fuels, cannot fail to avoid. (author)

  12. Innovation

    Science.gov (United States)

    EPA frames innovation as critical to the protection of human health and the environment through initiatives such as sustainable practices, innovative research, prize competitions, innovation awards, partnerships, and community activities.

  13. Operation management of the prototype heavy water reactor 'Fugen'

    International Nuclear Information System (INIS)

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

    1983-01-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

  14. The design features of integrated modular water reactor (IMR)

    International Nuclear Information System (INIS)

    Kanagawa, T.; Goto, M.; Usui, S.; Suzuta, T.; Serizawa, A.; Kunugi, T.; Yamauchi, T.; Itoh, G.; Matsumura, T.

    2004-01-01

    Small-to-medium-sized (300-600 MWe) reactors are required for the electric power market in the near future (2010-2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and small-scale design is needed. From such point of view, Integrated Modular Water Reactor (IMR), whose electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of the reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid systems, and Instrumentation and Control systems of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented. (authors)

  15. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  16. 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

  17. Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

    2009-01-01

    R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

  18. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  19. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  20. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same