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Sample records for cooled fbnr nuclear

  1. Sustainability and the Fixed Bed Nuclear Reactor (FBNR

    Directory of Open Access Journals (Sweden)

    Farhang Sefidvash

    2012-08-01

    Full Text Available Sustainability as a multifaceted and holistic concept is analyzed. Sustainability involves human relationship with elements such as natural environment, economy, power, governance, education and technology with the ultimate purpose of carrying forward an ever-advancing civilization. The Fixed Bed Nuclear Reactor (FBNR is an innovative, small, simple in design, inherently safe, non-proliferating, and environmentally friendly concept that its deployment can generate energy in a sustainable manner contributing to the prosperity of humanity. The development of FBNR will provide electricity as well as desalinated water through a simple but advanced technology for the developing, as well as developed countries. FBNR is environmentally friendly due to its inherent safety and the convenience of using its spent fuel as the source of radiation for irradiation purposes in agriculture, industry, and medicine. Politically, if a ping pong game brought peace between China and USA, a program of development of FBNR supported by the peace loving international community can become a more mature means to bring peace among certain apparently hostile nations who crave sustainable energy, desalinated water and simple advanced technology.

  2. Passive deca-heat removal in the fixed bed nuclear reactor (FBNR) - 15551

    International Nuclear Information System (INIS)

    Solano Diaz, E.C.; Luna Aguilera, G.M.; Santos, R.A.; Vaca, D.E.

    2015-01-01

    The Fixed Bed Nuclear Reactor (FBNR) is a Generation IV small reactor concept, where the spherical elements contain Triso-type microspheres with UO 2 , which serves as nuclear fuel. In the event that adverse operation conditions occur, the water pump is automatically shut off and the fuel pebbles fall back by gravity into the fuel chamber. Since the FBNR relies on passive security systems, the removal of the decay heat in the fuel chamber is achieved by contact with quiescent water. In the present paper, a mathematical simulation of the passive cooling of the system was conducted in SOLIDWORKS so as to obtain a temperature profile in the body during the decay heat removal process. Homogenization techniques were employed to smooth out spatial variations across the multiphase system and to derive expression for the effective thermophysical properties that are valid through the macroscopic entry (the chamber). The simulation showed that the chamber's temperature rose from 573 K to its maximum temperature, 1234 K, in the first hour. Afterwards, the temperature fluctuated, but stayed under 552 K. Since the temperature of the system was always kept under the value of the safety parameter (1200 C. degrees) the simulation confirmed that an effective passive cooling of the fuel chamber is indeed feasible. (authors)

  3. Preparation of human resources for future nuclear energy using FBNR as the instrument of learning

    International Nuclear Information System (INIS)

    Sefidvash, Farhang; Espinoza, Patricio; Guerrero, Victor Hugo

    2015-01-01

    An increasing number of developing countries are showing interest to become the emerging countries to nuclear energy. Most of these countries lack human resources and adequate infrastructures to enter such a venture. The principle objective of activities of FBNR Group is to train human resources for the countries that at the present lack the necessary conditions, but aim at the future clean and safe nuclear energy through the fourth generation and INPRO compatible nuclear reactors. The preparation for the future nuclear energy is done through development of innovative nuclear reactor that meets the INPRO philosophies and criteria. These countries may or may not have decided as yet to utilize nuclear energy, but are interested to gain a strong educational foundation for their future. The research and development of a small innovative nuclear reactor FBNR is used as the instrument for learning. The young scientists will learn how to be innovative with the vision of INPRO philosophy and criteria.

  4. Preparation of human resources for future nuclear energy using FBNR as the instrument of learning

    Energy Technology Data Exchange (ETDEWEB)

    Sefidvash, Farhang; Espinoza, Patricio; Guerrero, Victor Hugo [Escuela Politecnica Nacional (EPN), Quito (Ecuador); and others

    2015-11-15

    An increasing number of developing countries are showing interest to become the emerging countries to nuclear energy. Most of these countries lack human resources and adequate infrastructures to enter such a venture. The principle objective of activities of FBNR Group is to train human resources for the countries that at the present lack the necessary conditions, but aim at the future clean and safe nuclear energy through the fourth generation and INPRO compatible nuclear reactors. The preparation for the future nuclear energy is done through development of innovative nuclear reactor that meets the INPRO philosophies and criteria. These countries may or may not have decided as yet to utilize nuclear energy, but are interested to gain a strong educational foundation for their future. The research and development of a small innovative nuclear reactor FBNR is used as the instrument for learning. The young scientists will learn how to be innovative with the vision of INPRO philosophy and criteria.

  5. Passive cooling of a fixed bed nuclear reactor

    International Nuclear Information System (INIS)

    Petry, V.J.; Bortoli, A.L. de; Sefidwash, F.

    2005-01-01

    Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. Here the passive cooling characteristic of the fixed bed nuclear reactor (FBNR), that is being developed under the International Atomic Energy Agency (IAEA) Coordinated Research Project, is studied. A mathematical model is developed to calculate the temperature distribution in the fuel chamber of the reactor. The results demonstrate the passive cooling of this nuclear reactor concept. (authors)

  6. Thermal-hydraulic study of fixed bed nuclear reactor (FBNR), in FCC, BCC and pseudo-random configurations of the core through CFD method

    International Nuclear Information System (INIS)

    Luna, M.; Chavez, I.; Cajas, D.; Santos, R.

    2015-01-01

    The study of thermal-hydraulic performance of a fixed bed nuclear reactor (FBNR) core and the effect of the porosity was studied by the CFD method with 'SolidWorks' software. The representative sections of three different packed beds arrangements were analyzed: face-centered cubic (FCC), body-centered cubic (BCC), and a pseudo-random, with values of porosity of 0.28, 0.33 and 0.53 respectively. The minimum coolant flow required to avoid the phase change for each one of the configurations was determined. The results show that the heat transfer rate increases when the porosity value decreases, and consequently the minimum coolant flow in each configuration. The results of minimum coolant flow were: 728.51 kg/s for the FCC structure, 372.72 kg/s for the BCC, and 304.96 kg/s for the pseudo-random. Meanwhile, the heat transfer coefficients in each packed bed were 6480 W/m 2 *K, 3718 W/m 2 *K and 3042 W/m 2 *K respectively. Finally the pressure drop was calculated, and the results were 0.588 MPa for FCC configuration, 0.033 MPa for BCC and 0.017 MPa for the pseudo-random one. This means that with a higher porosity, the fluid can circulate easier because there are fewer obstacles to cross, so there are fewer energy losses. (authors)

  7. History of nuclear cooling

    International Nuclear Information System (INIS)

    Kuerti, M.

    1998-01-01

    The historical development of producing extreme low temperatures by magnetic techniques is overviewed. With electron spin methods, temperatures down to 1 mK can be achieved. With nuclear spins theoretically 10 -9 K can be produced. The idea of cooling with nuclear demagnetization is not new, it is a logical extension of the concept of electron cooling. Using nuclear demagnetization experiment with 3 T water cooled solenoids 3 mK could be produced. The cold record is held by Olli Lounasmaa in Helsinki with temperatures below 10 -9 K. (R.P.)

  8. 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

  9. Cooling nuclear reactor fuel

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1975-01-01

    Reference is made to water or water/steam cooled reactors of the fuel cluster type. In such reactors it is usual to mount the clusters in parallel spaced relationship so that coolant can pass freely between them, the coolant being passed axially from one end of the cluster in an upward direction through the cluster and being effective for cooling under normal circumstances. It has been suggested, however, that in addition to the main coolant flow an auxiliary coolant flow be provided so as to pass laterally into the cluster or be sprayed over the top of the cluster. This auxiliary supply may be continuously in use, or may be held in reserve for use in emergencies. Arrangements for providing this auxiliary cooling are described in detail. (U.K.)

  10. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  11. Nuclear reactor cooling device

    International Nuclear Information System (INIS)

    Hoshi, Masaya; Makihara, Yoshiaki.

    1985-01-01

    Purpose: To improve the heat transfer performance, as well as reducing and simplifying the structure while preventing the intrusion of primary coolants to utilization systems. Constitution: Heat transfer from the primary coolant circuit to the utilization circuits is conducted by means of heat pipe type heat exchangers. The heat exchanger comprises a tightly closed vessel divided by a partition wall, through which a plurality of heat pipes are passed. The primary coolants receiving the heat from the nuclear reactor enter the first chamber of the heat exchanger to heat the evaporating portion of the heat pipes. The heated flow of steams in the heat pipes transfer to the condensating portion in the second chamber to conduct heat exchange with the utilization system. In this way, since secondary coolant circuits are saved, the heat transfer performance can be improved significantly and the risk of failure can be reduced. (Kamimura, M,)

  12. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    1974-01-01

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

  13. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Guidez, Joel; Jarriand, Paul.

    1975-01-01

    The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

  14. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  15. Cooling towers of nuclear power plants

    International Nuclear Information System (INIS)

    Mikyska, L.

    1986-01-01

    The specifications are given of cooling towers of foreign nuclear power plants and a comparison is made with specifications of cooling towers with natural draught in Czechoslovak nuclear power plants. Shortcomings are pointed out in the design of cooling towers of Czechoslovak nuclear power plants which have been derived from conventional power plant design. The main differences are in the adjustment of the towers for winter operation and in the designed spray intensity. The comparison of selected parameters is expressed graphically. (J.B.)

  16. The atmospheric cooling of nuclear power stations

    International Nuclear Information System (INIS)

    Leuenberger, J.M.; Mayor, J.C.; Gassmann, F.; Lieber, K.

    1978-08-01

    Four different types of nuclear reactor are considered: light water reactors, high temperature reactors with steam circulation and with direct gas turbine circulation, and fast breeder reactors. Wet and dry cooling towers are described and experimental studies carried out using cooling tower models are presented. (G.T.H.)

  17. Helium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Longton, P.B.; Cowen, H.C.

    1975-01-01

    In helium cooled HTR's there is a by-pass circuit for cleaning purposes in addition to the main cooling circuit. This is to remove such impurities as hydrogen, methane, carbon monoxide and water from the coolant. In this system, part of the coolant successively flows first through an oxidation bed of copper oxide and an absorption bed of silica gel, then through activated charcoal or a molecular sieve. The hydrogen and carbon monoxide impurities are absorbed and the dry gas is returned to the main cooling circuit. To lower the hydrogen/water ratio without increasing the hydrogen fraction in the main cooling circuit, some of the hydrogen fraction converted into water is added to the cooling circuit. This is done, inter alia, by bypassing the water produced in the oxidation bed before it enters the absorption bed. The rest of the by-pass circuit, however, also includes an absorption bed with a molecular sieve. This absorbs the oxidized carbon monoxide fraction. In this way, such side effects as the formation of additional methane, carburization of the materials of the by-pass circuit or loss of graphite are avoided. (DG/RF) [de

  18. Cooling water requirements and nuclear power plants

    International Nuclear Information System (INIS)

    Rao, T.S.

    2010-01-01

    Indian nuclear power programme is poised to scuttle the energy crisis of our time by proposing joint ventures for large power plants. Large fossil/nuclear power plants (NPPs) rely upon water for cooling and are therefore located near coastal areas. The amount of water a power station uses and consumes depends on the cooling technology used. Depending on the cooling technology utilized, per megawatt existing NPPs use and consume more water (by a factor of 1.25) than power stations using other fuel sources. In this context the distinction between 'use' and 'consume' of water is important. All power stations do consume some of the water they use; this is generally lost as evaporation. Cooling systems are basically of two types; Closed cycle and Once-through, of the two systems, the closed cycle uses about 2-3% of the water volumes used by the once-through system. Generally, water used for power plant cooling is chemically altered for purposes of extending the useful life of equipment and to ensure efficient operation. The used chemicals effluent will be added to the cooling water discharge. Thus water quality impacts on power plants vary significantly, from one electricity generating technology to another. In light of massive expansion of nuclear power programme there is a need to develop new ecofriendly cooling water technologies. Seawater cooling towers (SCT) could be a viable option for power plants. SCTs can be utilized with the proper selection of materials, coatings and can achieve long service life. Among the concerns raised about the development of a nuclear power industry, the amount of water consumed by nuclear power plants compared with other power stations is of relevance in light of the warming surface seawater temperatures. A 1000 MW power plant uses per day ∼800 ML/MW in once through cooling system; while SCT use 27 ML/MW. With the advent of new marine materials and concrete compositions SCT can be constructed for efficient operation. However, the

  19. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  20. Nuclear demagnetisation cooling of a nanoelectronic device

    Science.gov (United States)

    Jones, Alex; Bradley, Ian; Guénault, Tony; Gunnarsson, David; Haley, Richard; Holt, Stephen; Pashkin, Yuri; Penttilä, Jari; Prance, Jonathan; Prunnila, Mika; Roschier, Leif

    We present a new technique for on-chip cooling of electrons in a nanostructure: nuclear demagnetisation of on-chip, thin-film copper refrigerant. We are motivated by the potential improvement in the operation of nanoelectronic devices below 10 mK . At these temperatures, weak electron-phonon coupling hinders traditional cooling, yet here gives the advantage of thermal isolation between the environment and the on-chip electrons, enabling cooling significantly below the base temperature of the host lattice. To demonstrate this we electroplate copper onto the metallic islands of a Coulomb blockade thermometer (CBT), and hence provide a direct thermal link between the cooled copper nuclei and the device electrons. The CBT provides primary thermometry of its internal electron temperature, and we use this to monitor the cooling. Using an optimised demagnetisation profile we observe the electrons being cooled from 9 mK to 4 . 5 mK , and remaining below 5 mK for an experimentally useful time of 1200 seconds. We also suggest how this technique can be used to achieve sub- 1 mK electron temperatures without the use of elaborate bulk demagnetisation stages.

  1. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  2. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  3. Model of cooling nuclear fuel rod in the nuclear reactor

    International Nuclear Information System (INIS)

    Lavicka, David; Polansky, Jiri

    2010-01-01

    The following topics are described: Some basic requirements for nuclear fuel rods; The VVER 1000 fuel rod; Classification of the two-phase flow in the vertical tube; Type of heat transfer crisis in the vertical tube; Experimental apparatus; Model of the nuclear fuel rod and spacers; Potential of the experimental apparatus (velocity profile measurement via PIV; thermal flow field measurement by the PLIF method; cooling graph in dependence on the fuel rod temperature; comparison of the hydrodynamic properties with respect to the design features of the spacers). (P.A.)

  4. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

  5. 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

  6. Cooling facility of nuclear power plant

    International Nuclear Information System (INIS)

    Arai, Kenji; Nagasaki, Hideo.

    1992-01-01

    In a cooling device of a nuclear power plant, an exhaust pipe for an incondensible gas is branched. One of the branched exhaust pipes is opened in a pressure suppression pool water in a suppression chamber containing pool water and the other is opened at a lower portion of a dry well incorporating a pressure vessel. In a state where the pressure in the dry well is higher than that in the suppression chamber, an off-gas is exhausted effectively by way of the exhaustion pipe in communication with the suppression chamber. In a state where there is no difference between the pressures and the opening end of the exhaustion pipe in communication with the suppression chamber is sealed with water, off-gas is exhausted by way of the exhaustion pipe in communication with the lower portion of the dry well. Then, since the incondensible gas in a heat transfer pipe is not accumulated, after-heat can be removed efficiently. Satisfactory cooling is maintained even after the coincidence of the pressures in the dry well with that in the suppression chamber, to decrease a pressure in a reactor container. (N.H.)

  7. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  8. Cooling water facilities at a nuclear station

    International Nuclear Information System (INIS)

    Hurst, W.L.; Ghadiali, B.M.; Kanovich, J.S.

    1983-01-01

    The use of ponds for holding a reserve of cooling water obtained as sewage effluent and also for collection of waste water for disposal by evaporation, was made at a nuclear power plant site in southern Arizona. The power output of the plant will be 3,900 MW. Two single cell ponds are 80 acres (30 ha) and 250 acres (100 ha) in size. Excavated materials from the 80-acre (30ha) pond were used for structural backfill as planned, and the 250-acre (100ha) pond was designed for limited dike height with balanced cut and fill and some excess materials used as side berms for additional safety. Both ponds are being lined with a unique combination of linings to provide environmental safeguards and at the same time cost-effectiveness is compared to alternative schemes

  9. Corrosion by cooling gases in nuclear reactors

    International Nuclear Information System (INIS)

    Darras, R.

    1960-01-01

    This article begins with a review of the various materials which can be used and the cooling gases in which they may be heated, emphasis being placed on the importance of reaching temperatures as high as possible. This is followed by a few general remarks on the dry oxidation of metals and alloys, particularly with regard to diffusion phenomena and their various possible mechanisms, and also the methods of investigation employed. Finally, the behaviour of the chief nuclear materials heated in the various gases is studied successively. Materials used for fuel (metallic uranium, uranium oxide, carbides and silicides), canning materials (magnesium, aluminium, zirconium, beryllium, stainless and refractory steels), structural materials (ordinary or slightly alloyed steels), and finally moderators (graphite, beryllium oxide) are deal with in this way. This account is backed up both by the results obtained at the CEA and by work published outside or abroad up to the present day. In conclusion, every effort has been made to direct future research on the basis of the foregoing. Reprint of a paper published in Industries Atomiques - no. 9/10, 1959, p. 3-23 [fr

  10. 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

  11. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  12. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

  13. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

    Rouai, N. M.

    1998-01-01

    This paper presents some recent experimental results performed with the aim of understanding the mechanism of passive cooling. The AP 600 passive containment cooling system is simulated by an electrically heated vertical pipe, which is cooled by a naturally induced air flow and by a water film descending under gravity. The results demonstrate that although the presence of the water film improved the heat transfer significantly, the mode of heat transfer was very dependent on the experimental parameters. Preheating the water improved both film stability and overall cooling performance

  14. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  15. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  16. Closed-cycle cooling systems for nuclear power plants

    International Nuclear Information System (INIS)

    Santini, Lorenzo

    2006-01-01

    The long experience in the field of closed-cycle cooling systems and high technological level of turbo machines and heat exchangers concurs to believe in the industrial realizability of nuclear systems of high thermodynamic efficiency and intrinsic safety [it

  17. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  18. Kaiseraugst nuclear power station: meteorological effects of the cooling towers

    International Nuclear Information System (INIS)

    Anon.

    1975-01-01

    Considerations of water conservation persuaded the German Government in 1971 not to allow the use of the Aar and Rhine for direct cooling of nuclear power stations. The criticism is often made that the Kaiseraugst cooling towers were built without full consideration of the resulting meteorological effects. The criticism is considered unjustified because the Federal Cooling Tower Commission considered all the relevant aspects before making its recommendations in 1972. Test results and other considerations show that the effect of the kaiseraugst cooling towers on meteorological and climatic conditions is indeed minimal and details are given. (P.G.R.)

  19. Nuclear reactor lid cooling which can work by natural circulation

    International Nuclear Information System (INIS)

    Wagner, J.

    1985-01-01

    The well-known air cooling of the lid of liquid metal cooled nuclear reactors is improved by the start of natural convection flow ensuring removal of heat in a sufficiently short time, if the blower fails. Go and return branches of the individual cooling circuits are arranged at different heights for this purpose. The circulation is supported by opening valves, which provide a direct path into the reactor building for the cooling air. The draught can be increased by setting up special chimneys. The start of circulation is aided by the temporary opening of another valve. (orig.) [de

  20. Cooling system for auxiliary systems of a nuclear power plant

    International Nuclear Information System (INIS)

    Maerker, W.; Mueller, K.; Roller, W.

    1981-01-01

    From the reactor auxiliary and ancillary systems of a nuclear facility heat has to be removed without the hazard arising that radioactive liquids or gases may escape from the safe area of the nuclear facility. A cooling system is described allowing at every moment to make available cooling fluid at a temperature sufficiently low for heat exchangers to be able to remove the heat from such auxiliary systems without needing fresh water supply or water reservoirs. For this purpose a dry cooling tower is connected in series with a heat exchanger that is cooled on the secondary side by means of a refrigerating machine. The cooling pipes are filled with a nonfreezable fluid. By means of a bypass a minimum temperature is guaranteed at cold weather. (orig.) [de

  1. Cooling water recipients for nuclear power plants

    International Nuclear Information System (INIS)

    Dahl, F.-E.; Saetre, H.J.

    1971-10-01

    The hydrographical and hydrological conditions at 17 prospective nuclear power plant sites in the Oslofjord district are evaluated with respect to their suitability as recipients for thermal discharges from nuclear power plants. No comparative evaluations are made. (JIW)

  2. Emergency core cooling systems in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1981-12-01

    This report contains the responses by the Advisory Committee on Nuclear Safety to three questions posed by the Atomic Energy Control Board concerning the need for Emergency Core Cooling Systems (ECCS) in CANDU nuclear power plants, the effectiveness requirement for such systems, and the extent to which experimental evidence should be available to demonstrate compliance with effectiveness standards

  3. Kaon condensates, nuclear symmetry energy and cooling of neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Kubis, S. E-mail: kubis@alf.ifj.edu.pl; Kutschera, M

    2003-06-02

    The cooling of neutron stars by URCA processes in the kaon-condensed neutron star matter for various forms of nuclear symmetry energy is investigated. The kaon-nucleon interactions are described by a chiral Lagrangian. Nuclear matter energy is parametrized in terms of the isoscalar contribution and the nuclear symmetry energy in the isovector sector. High density behaviour of nuclear symmetry energy plays an essential role in determining the composition of the kaon-condensed neutron star matter which in turn affects the cooling properties. We find that the symmetry energy which decreases at higher densities makes the kaon-condensed neutron star matter fully protonized. This effect inhibits strongly direct URCA processes resulting in slower cooling of neutron stars as only kaon-induced URCA cycles are present. In contrast, for increasing symmetry energy direct URCA processes are allowed in the almost whole density range where the kaon condensation exists.

  4. Kaon condensates, nuclear symmetry energy and cooling of neutron stars

    International Nuclear Information System (INIS)

    Kubis, S.; Kutschera, M.

    2003-01-01

    The cooling of neutron stars by URCA processes in the kaon-condensed neutron star matter for various forms of nuclear symmetry energy is investigated. The kaon-nucleon interactions are described by a chiral Lagrangian. Nuclear matter energy is parametrized in terms of the isoscalar contribution and the nuclear symmetry energy in the isovector sector. High density behaviour of nuclear symmetry energy plays an essential role in determining the composition of the kaon-condensed neutron star matter which in turn affects the cooling properties. We find that the symmetry energy which decreases at higher densities makes the kaon-condensed neutron star matter fully protonized. This effect inhibits strongly direct URCA processes resulting in slower cooling of neutron stars as only kaon-induced URCA cycles are present. In contrast, for increasing symmetry energy direct URCA processes are allowed in the almost whole density range where the kaon condensation exists

  5. 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

  6. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  7. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  8. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  9. Thermodynamic analysis of cooling systems for nuclear power stations condenser

    International Nuclear Information System (INIS)

    Beck, A.

    1985-06-01

    This work is an attempt to concentrate on the thermodynamic theory, the engineering solution and the quantities of water needed for the operation of a wet as well as a wet/dry cooling towers coupled to a nuclear turbine condenser,. About two hundred variables are needed for the design of a condenser - cooling tower system. In order to make the solution fast and handy, a computer model was developed. The amount of water evaporation from cooling towers is a function of the climate conditions prevailing around the site. To achieve an authentic analysis, the meteorological data of the northern Negev was used. The total amount of water necessary to add to the system in a year time of operation is large and is a function of both the blow-down rate and the evaporation. First estimations show that the use of a combined system, wet/dry cooling tower, is beneficial in the northern Negev area. Such a system can reduce significantly the amount of wasted fresh water. Lack of international experience is the major problem in the acceptability of wet/dry cooling towers. The technology of a wet cooling tower using sea water is also discussed where no technical or engineering limitations were found. This work is an attempt to give some handy tools for making the choice of cooling systems for nuclear power plants easier

  10. Status of helium-cooled nuclear power systems. [Development potential

    Energy Technology Data Exchange (ETDEWEB)

    Melese-d' Hospital, G.; Simnad, M

    1977-09-01

    Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor is also discussed. The status of helium-cooled nuclear energy systems is summarized as a basis for assessing their prospects. 50 references.

  11. Lower parts of Temelin nuclear power plant cooling towers

    International Nuclear Information System (INIS)

    Sebek, J.

    1988-01-01

    The progress of work is described in detail on the foundations and lower parts of the cooling towers of the Temelin nuclear power plant. The cooling tower is placed on a reinforced concrete footing of a circular layout. Support pillars are erected on the reinforced concrete continuous footing. They consists of oblique shell stanchions. Inside, the footing joins up to monolithic wall and slab structures of the cooling tower tub. The tub bottom forms a foundation plate supporting prefab structures of the cooling tower inner structural systems. The framed support of the chimney shell consists of 56 pairs of prefabricated oblique stanchions. Following their erection into the final position and anchoring in the continuous footing, the concreting of the casing can start of the reinforced conrete chimney. (Z.M.). 3 figs

  12. Water cooled type nuclear power plant

    International Nuclear Information System (INIS)

    Arai, Shigeki.

    1981-01-01

    Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

  13. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

  14. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  15. Secondary process for securing emergency cooling in nuclear facilities

    International Nuclear Information System (INIS)

    Bachl, H.

    1975-01-01

    An auxiliary process for securing the emergency cooling of nuclear power plants is described which is characterized in that a two-material heat power auxiliary process is connected at the cold end of the cooling circuit to a main heat power process to obtain mechanical energy from thermal, which in normal operation works as a cold-absorption process, but with failure of the main process changes to a heat power process with full evaporation and subsequent superheating of the two-materials mixture. (RW/LH) [de

  16. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  17. Advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2008-07-01

    By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average nuclear plant availability during 2006 reached 83%; - Growing energy needs: All forecasts project increases in world energy demand, especially as population and economic productivity grow. The strategies are country dependent, but usually involve a mix of energy sources; - Interest in advanced applications of nuclear energy, such as seawater desalination, steam for heavy oil recovery and heat and electricity for hydrogen production; - Environmental concerns and constraints: The Kyoto Protocol has been in force since February 2005, and for many countries (most OECD countries, the Russian Federation, the Baltics and some countries of the Former Soviet Union and Eastern Europe) greenhouse gas emission limits are imposed; - Security of energy supply is a national priority in essentially every country; and - Nuclear power is economically competitive and provides stability of electricity price. In the near term most new nuclear plants will be evolutionary water cooled reactors (Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs), often pursuing economies of scale. In the longer term, innovative designs that

  18. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  19. Cooling and heating facility for nuclear power plant

    International Nuclear Information System (INIS)

    Kakuta, Atsuro

    1994-01-01

    The present invention concerns a cooling and heating facility for a nuclear power plant. Namely, a cooling water supply system supplies cooling water prepared by a refrigerator for cooling the inside of the plant. A warm water supply system supplies warm water having its temperature elevated by using an exhausted heat from a reactor water cleanup system. The facility comprises a heat pump-type refrigerator disposed in a cold water supply system for producing cold water and warm water, and warm water pipelines for connecting the refrigerator and the warm water supply system. With such a constitution, when the exhaust heat from the reactor water cleanup system can not be used, warm water prepared by the heat pump type refrigerator is supplied to the warm water supply system by way of the warm water pipelines. Accordingly, when the exhaust heat from the reactor water cleanup system can not be used such as upon inspection of the plant, a portion of the refrigerators in a not-operated state can be used for heating. Supply of boiler steams in the plant is no more necessary or extremely reduced. (I.S.)

  20. Fuel rod for liquid metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vinz, P.

    1976-01-01

    In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

  1. Indian programme on molten salt cooled nuclear reactors

    International Nuclear Information System (INIS)

    DuIera, I.V.; Vijayan, P.K.; Sinha, R.K.

    2013-01-01

    Bhabha Atomic Research Centre (BARC) is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of molten fluoride salts and is capable of supplying process heat at 1000 ℃ to facilitate hydrogen production by splitting water. BARC has also initiated studies for a reactor concept in which salts of molten fluoride fuel and coolant in fluid form, flows through the reactor core of graphite moderator, resulting in nuclear fission within the molten salt. For thorium fuel cycle, this concept is very attractive, since the fuel can be re-processed on-line, enabling it to be an efficient neutron breeder. (author)

  2. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  3. 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)

  4. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    International Nuclear Information System (INIS)

    Hunsbedt, A.; Boardman, C.E.

    1993-01-01

    A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

  5. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  6. 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)

  7. 77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

    Science.gov (United States)

    2012-12-07

    ... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

  8. Visualization test facility of nuclear fuel rod emergency cooling system

    International Nuclear Information System (INIS)

    Candido, Marcos Antonio; Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole

    2013-01-01

    The nuclear reactors safety is determined according to their protection against the consequences that may result from postulated accidents. The Loss of Coolant Accident (LOCA) is one the most important design basis accidents (DBA). The failure may be due to rupture of the primary loop piping. Another accident postulated is due to lack of power in the pump motors in the primary circuit. In both cases the reactor shut down automatically due to the decrease of reactivity to maintain the fissions, and to the drop of control rods. In the event of an accident it is necessary to maintain the coolant flow to remove the fuel elements residual heat, which remains after shut down. This heat is a significant amount of the maximum thermal power generated in normal operation (about 7%). Recently this event has been quite prominent in the press due to the reactor accident in Fukushima nuclear power station. This paper presents the experimental facility under rebuilding at the Thermal Hydraulic Laboratory of the Nuclear Technology Development Center (CDTN) that has the objective of monitoring and visualization of the process of emergency cooling of a nuclear fuel rod simulator, heated by Joule effect. The system will help the comprehension of the heat transfer process during reflooding after a loss of coolant accident in the fuel of light water reactor core. (author)

  9. Tests of cooling water pumps at Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Travnicek, J.

    1986-01-01

    Tests were performed to examine the operating conditions of the 1600 BQDV cooling pumps of the main coolant circuit of unit 1 of the Dukovany nuclear power plant. For the pumps, the performance was tested in the permissible operating range, points were measured below this range and the guaranteed operating point was verified. Pump efficiency was calculated from the measured values. The discussion of the measurement of parameters has not yet been finished because the obtained values of the amount delivered and thus of the pump efficiency were not up to expectation in all detail. It was also found that for obtaining the guaranteed flow the pump impeller had to be opened to 5deg -5.5deg instead of the declared 3deg. Also tested were pump transients, including the start of the pump, its stop, the operation and failure of one of the two pumps. In these tests, pressures were also measured at the inlet and the outlet of the inner part of the TG 11 turbine condenser. It was shown that the time course and the pressure course of the processes were acceptable. In addition to these tests, pressure losses in the condenser and the cooling water flow through the feed pump electromotor cooler wre tested for the case of a failure of one of the two pumps. (E.S.)

  10. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

  11. Acoustical environment of gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Blevins, R.D.

    1986-01-01

    Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

  12. High temperature friction and seizure in gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Cousseran, P.; Febvre, A.; Martin, R.; Roche, R.

    1978-01-01

    One of the most delicate problems encountered in the gas cooled nuclear reactors is the friction without lubrication in a dry and hot (800 0 C /1472 0 F) helium atmosphere even at very small velocity. The research and development programs are described together with special tribometers that operate at mode than 1000 0 C (1832 0 F) in dry helium. The most interesting test conditions and results are given for gas nitrited steels and for strongly alloyed Ni-Cr steels coated with chromium carbide by plasma sprayed. The effects of parameters live velocity, travelled distance, contact pressure, roughness, temperature and prolonged stops under charge are described together with the effects of negative phenomena like attachment and chattering [fr

  13. Numerical modeling of a nuclear production reactor cooling lake

    International Nuclear Information System (INIS)

    Hamm, L.L.; Pepper, D.W.

    1987-01-01

    A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation

  14. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  15. Materials for innovative lead alloy cooled nuclear systems: Overview

    International Nuclear Information System (INIS)

    Mueller, Georg; Weisenburger, Alfons; Fetzer, Renate; Heinzel, Annette; Jianu, Adrian

    2015-01-01

    One of the most challenging issues for all future innovative nuclear systems including Gen IV reactors are materials. The selection of the structural materials determines the design which has to consider the properties and the availability of the materials. Beside general requirements for material properties that are common for all fast reactor types specific issues arise from coolant compatibility. The high solubility of steel alloying elements in liquid Pb-alloys at reactor relevant temperatures is clearly detrimental. Therefore, all steels that are considered as structural materials have to be protected by dissolution barriers. The most common barriers for steels under consideration are oxide scales that form in situ during operation. However, increasing the temperature above 500 deg. C will result either in dissolution attack or in enhanced oxidation. For higher temperatures additional barriers like alumina forming surface alloys are discussed and investigated. Mechanical loads like creep stress and fretting will act on the steels. These mechanical loads will interact with the coolant and can increase the negative effects. For a LFR (Lead Fast Reactor) Demonstrator and MYHRRA (ADS) austenitic steels (316L) are selected for most in core components. The 15-15Ti is the choice for the fuel cladding of MYHRRA and a Pb cooled demonstrator. For an industrial LFR (Lead Fast Reactor) the ferritic martensitic steel T91 was selected as fuel clad material due to its improved irradiation resistance. T91 is in both designs the material to be used for the heat exchanger. Surface alloying with alumina forming alloys is considered to assure material functionality at higher temperatures and is therefore selected for fuel cladding of the ELFR and the heat exchanger tubes. This presentation will give an overview on the selected materials for innovative Pb alloy cooled nuclear systems considering, beside pure compatibility, the influence of mechanical interaction like creep and

  16. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  17. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    Science.gov (United States)

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  18. Hydrogen in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The Commission of the European Community (CEC) and the International Atomic Energy Agency (IAEA) decided in 1989 to update the state of the art concerning hydrogen in water cooled nuclear power reactors by commissioning a report which would review, all the available information to-date and make recommendations for the future. This joint report was prepared by committees formed by the IAEA and by the CEC. The aim of this report is to review the current understanding on the areas in which the research on hydrogen in LWR is conventionally presented, taking into account the results of the latest reported research developments. The main reactions through which hydrogen is produced are assessed together with their timings. An estimation of the amount of hydrogen produced by each reaction is given, in order to reckon their relative contribution to the hazard. An overview is then given of the state of knowledge of the most important phenomena taking place during its transport from the place of production and the phenomena which control the hydrogen combustion and the consequences of combustion under various conditions. Specific research work is recommended in each sector of the presented phenomena. The last topics reviewed in this report are the hydrogen detection and the prevent/mitigation of pressure and temperature loads on containment structures and structures and safety related equipment caused by hydrogen combustion

  19. A dilution refrigerator combining low base temperature, high cooling power and low heat leak for use with nuclear cooling

    International Nuclear Information System (INIS)

    Bradley, D.I.; Guenault, A.M.; Keith, V.; Miller, I.E.; Pickett, G.R.; Bradshaw, T.W.; Locke-Scobie, B.G.

    1982-01-01

    The design philosophy, design, construction and performance of a dilution refrigerator specifically intended for nuclear cooling experiments in the submillikelvin regime is described. Attention has been paid from the outset to minimizing sources of heat leaks, and to achieving a low base temperature and relatively high cooling power below 10 mK. The refrigerator uses sintered silver heat exchangers similar to those developed at Grenoble. The machine has a base temperature of 3 mK or lower and can precool a copper nuclear specimen in 6.8 T to 8 mK in 70 h. The heat leak to the innermost nuclear stage is < 30 pW after only a few days' running. (author)

  20. Sodium leak detection system for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Modarres, D.

    1991-01-01

    This patent describes a device for detecting sodium leaks from a reactor vessel of a liquid sodium cooled nuclear reactor the reactor vessel being concentrically surrounded by a a containment vessel so as to define an airtight gap containing argon. It comprises: a light source for generating a first light beam, the first light beam having first and second predominant wavelengths, the first wavelength being substantially equal to an absorption line of sodium and the second wavelength being chosen such that it is not absorbed by sodium and argon; an optical multiplexer optically coupled to the light source; optically coupled to the multiplexer, each of the sensors being embedded in the containment vessel of the reactor, each of the sensors projecting the first light beam into the gap and collecting the first light beam after it has reflected off of a surface of the reactor vessel; a beam splitter optically coupled to each of the sensors through the multiplexer, the beam splitter splitting the first light beam into second and third light beams of substantially equal intensities; a first filter dispersed within a path of second light beam for filtering the second wavelength out of the third light beam; first and second detector beams disposed with in the paths of the second and third light beams so as to detect the intensities of the second and third light beams, respectively; and processing means connected to the first and second detector means for calculating the amount of the first wavelength which is absorbed when passing through the argon

  1. Emergency cooling process and device for nuclear reactor containment

    International Nuclear Information System (INIS)

    Costes, D.

    1985-01-01

    The emergency cooling system of a PWR containment, according to the principal patent, comprises a turbine fed by the humid air of the containment, a condenser in which the air flowing out of the turbine is dryed and cooled by an external coolant and a compressor actuated by the turbine and returning the dryed air in the containment [fr

  2. Emergency cooling method and system for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1982-01-01

    For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

  3. State of the art of nuclear facilities with organic cooled reactors

    International Nuclear Information System (INIS)

    Brede, O.

    1984-01-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

  4. Fracture during cooling of cast borosilicate glass containing nuclear wastes

    International Nuclear Information System (INIS)

    Smith, P.K.; Baxter, C.A.

    1981-09-01

    Procedures and techniques were evaluated to mitigate thermal stress fracture in waste glass as the glass cools after casting. The two principal causes of fracture identified in small-scale testing are internal thermal stresses arising from excessive thermal gradients when cooled too fast, and shear fracturing in the surface of the glass because the stainless steel canister shrinks faster than the glass on cooling. Acoustic emission and ceramographic techniques were used to outline an annealing schedule that requires at least three weeks of controlled cooling below 550 0 C to avoid excessive thermal gradients and corresponding stresses. Fracture arising from canister interactions cannot be relieved by slow cooling, but can be eliminated for stainless steel canisters by using ceramic paper, ceramic or graphite paste linings, or by choosing a canister material with a thermal expansion coefficient comparable to, or less than, that of the glass

  5. Cost comparison of dry-type and conventional cooling systems for representative nuclear generating plans

    International Nuclear Information System (INIS)

    Rossie, J.P.; Cecil, E.A.; Young, R.O.

    1974-01-01

    Results are presented of studies comparing the use of dry-type cooling towers with conventional cooling methods for representative pressurized-water-reactor nuclear power plants. The studies were based on the hypothetical use of dry-type cooling towers for three nuclear power plants now under construction which were designed and are being built to use conventional cooling methods. One of the plants is located in the northeastern United States, one in the Southeast and one in the West. The report also presents the results of comparisons based on a hypothetical plant at a typical eastern United States site. The three electric utilities which participated in these studies have furnished actual construction cost information for the conventional cooling systems being constructed, and the authors have made construction estimates for economically optimum dry cooling systems which might have been built in place of the conventional cooling systems being constructed. The report compares the physical and operating characteristics of dry-type and conventional cooling systems as well as the relative economics of the different cooling methods. The effect of dry cooling on the bus-bar cost of power has been computed for the three selected plants and for the typical eastern plant

  6. Organohalogens in chlorinated cooling waters discharged from nuclear power stations

    International Nuclear Information System (INIS)

    Bean, R.M.; Mann, D.C.; Neitzel, D.A.

    1983-01-01

    For the power plant discharges studied to date, measured concentrations of trihalomethanes are lower than might be expected, particularly in cooling tower water, which can lose THMs to the atmosphere. In the cooling towers, where chlorine was added in higher concentrations and for longer residence times, halogenated phenols can contribute significantly to the total organic halogen content of the discharge. The way in which cooling towers are operated may also influence the production of halogenated phenols because they concentrate the incoming water by a factor of 4 or 5. In addition, the phenols, which act as a substrate for the halogenating agent, are also probably concentrated by the cooling tower operation and may be prevented from being biodegraded by addition of the same biocide that produces the halogenated phenols. 8 references, 4 tables

  7. Water cooling system for sintering furnaces of nuclear fuel pellets

    International Nuclear Information System (INIS)

    1996-01-01

    This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author)

  8. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  9. Cooling water in the study of nuclear power plants sites

    International Nuclear Information System (INIS)

    Martinez, J.J.C.

    1990-01-01

    The location of an electric power plant has its limitations as regards the availability of apt sites. The radiosanitary risk, seismic risk and the overload capacity of the ground can be generically enumerated, being the cooling water availability for an electric power plant a basic requirement. Diverse cooling systems may be employed but the aim must always be that thermal contamination in the immediate environment be the least possible. (Author) [es

  10. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  11. COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

    Energy Technology Data Exchange (ETDEWEB)

    Gary Vine

    2010-12-01

    This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

  12. COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

    International Nuclear Information System (INIS)

    Vine, Gary

    2010-01-01

    This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes 'Best Technology Available' for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant's steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R and D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

  13. IAEA'S study on advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Cleveland, J.; McDonald, A.; Rao, A.; )

    2008-01-01

    About one-fifth of the world's energy consumption is used for electricity generation, with nuclear power contributing approximately 15.2% of this electricity. However; most of the world's energy consumption is for heat and transportation. Nuclear energy has considerable potential to penetrate these energy sectors now served by fossil fuels that are characterized by price volatility and finite supply. Advanced applications of nuclear energy include seawater desalination, district heating, and heat for industrial processes. Nuclear energy also has potential to provide a near-term, greenhouse gas free, source of energy for transportation. These applications rely on a source of heat and electricity. Nuclear energy from water-cooled reactors, of course, is not unique in this sense. Indeed, higher temperature heat can be produced by burning natural gas and coal, or through the use of other nuclear technologies such as gas-cooled or liquid-metal-cooled reactors. Water-cooled reactors, however; are being deployed today while other reactor types have had considerably less operational and regulatory experience and will take still some time to be widely accepted in the market. Both seawater desalination and district heating with nuclear energy are well proven, and new seawater desalination projects using water-cooled reactors will soon be commissioned. Provision of process heat with nuclear energy can result in less dependence on fossil fuels and contribute to reductions of greenhouse gases. Importantly, because nuclear power produces base-load electricity at stable and predictable prices, it provides a greenhouse gas free source of electricity for transportation systems (trains and subways), and for electric and plug-in hybrid vehicles, and in the longer term nuclear energy could produce hydrogen for fuel cell vehicles, as well as for other components of a hydrogen economy. These advanced applications can play an important role in enhancing public acceptance of nuclear

  14. 78 FR 35330 - Initial Test Programs for Water-Cooled Nuclear Power Plants

    Science.gov (United States)

    2013-06-12

    ... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance. SUMMARY: The U.S... Programs for Water-Cooled Nuclear Power Plants.'' This guide describes the general scope and depth that the... power plants. ADDRESSES: Please refer to Docket ID NRC-2012-0293 about the availability of information...

  15. Design guide for heat transfer equipment in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    1975-07-01

    Information pertaining to design methods, material selection, fabrication, quality assurance, and performance tests for heat transfer equipment in water-cooled nuclear reactor systems is given in this design guide. This information is intended to assist those concerned with the design, specification, and evaluation of heat transfer equipment for nuclear service and the systems in which this equipment is required. (U.S.)

  16. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  17. Startup of Pumping Units in Process Water Supplies with Cooling Towers at Thermal and Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Berlin, V. V., E-mail: vberlin@rinet.ru; Murav’ev, O. A., E-mail: muraviov1954@mail.ru; Golubev, A. V., E-mail: electronik@inbox.ru [National Research University “Moscow State University of Civil Engineering,” (Russian Federation)

    2017-03-15

    Aspects of the startup of pumping units in the cooling and process water supply systems for thermal and nuclear power plants with cooling towers, the startup stages, and the limits imposed on the extreme parameters during transients are discussed.

  18. 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

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  19. Absorber rod driving into a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Elter, C.; Schmitt, H.; Schoening, J.

    1987-01-01

    The absorber rod consists of a hollow cylinder which has a layer of absorber material applied on its inside circumferential surface. The absorber rod is held via a guide sleeve, which is supported centrally in a hole in the side reflector. The guidance within the sleeve is provided by flanges on the hollow cylinder. The movement of the hollow cylinder is carried out hydraulically or pneumatically. A flow of cooling gas is used for cooling, which is passed through the inner central areas of the hollow cylinder and the guide sleeve. (DG) [de

  20. 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)

  1. Nuclear γ-ray spectroscopy of cool free atoms

    International Nuclear Information System (INIS)

    Rivlin, Lev A

    1999-01-01

    Consideration is given to the capabilities of gamma-ray spectroscopy of the nuclei of free neutral atoms cooled employing modern laser light-pressure techniques. This spectroscopy is comparable with the Mossbauer spectroscopy in respect of the expected resolving power. (laser applications and other topics in quantum electronics)

  2. The modular high-temperature gas-cooled reactor: A cost/risk competitive nuclear option

    International Nuclear Information System (INIS)

    Gotschall, H.L.

    1994-01-01

    The business risks of nuclear plant ownership are identified as a constraint on the expanded use of nuclear power. Such risks stem from the exacting demands placed on owner/operator organizations of current plants to demonstrate ongoing compliance with safety regulations and the resulting high costs for operation and maintenance. This paper describes the Modular High-Temperature Gas-Cooled Reactor (MHTGR) design, competitive economics, and approach to reducing the business risks of nuclear plant ownership

  3. Immersion cooling of silicon photomultipliers (SiPM) for nuclear medicine imaging applications

    International Nuclear Information System (INIS)

    Raylman, R.R.; Stolin, A.V.

    2016-01-01

    Silicon photomultipliers (SiPM) are compact, high amplification light detection devices that have recently been incorporated into magnetic field-compatible positron emission tomography (PET) scanners. To take full advantage of these devices, it is preferable to cool them below room temperature. Most current methods are limited to the cooling of individual detector modules, increasing complexity and cost of scanners made-up of a large number of modules. In this work we investigated a new method of cooling, immersion of the detector modules in non-electrically conductive, cooled liquid. A small-scale prototype system was constructed to cool a relatively large area SiPM-based, scintillator detector module by immersing it in a circulating bath of mineral oil. Testing demonstrated that the system rapidly decreased and stabilized the temperature of the device. Operation of the detector illustrated the expected benefits of cooling, with no apparent degradation of performance attributable to immersion in fluid. - Highlights: • Immersion cooling is new, simple and inexpensive method to cool solid state based nuclear medicine scanner. • Method successfully tested on a scaled version of an SiPM-based PET detector module. • Can be scaled up to cool a complete PET scanner.

  4. Forced cooling of a nuclear waste repository mine drift - a scoping analysis

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, R D [Sandia National Labs., Albuquerque, NM (USA)

    1982-12-01

    Nuclear waste repositories, with decay heat generation beneath the mine drift floors, are force-cooled with air so that re-entrance is possible many years after the waste has been buried. A numerical model has been developed which uses heat transfer coefficients as input. It has been demonstrated that mixed (forced and free) convective and surface roughness effects are significant and must be included in future experiments if reliable predictions are to be made of the time required to cool the repository. For example, when repository mine drifts in volcanic tuff are force-cooled, with forced convection being the only energy transport mechanism, it takes approx.= 0.1 year to cool the mine surface to a safe temperature. However, when mixed convection is the primary transport mechanism it takes approx.= 1.0 year to cool the mine. In addition to mixed convection, other effects are delineated.

  5. Environmental impact of cooling towers of large nuclear power plants

    International Nuclear Information System (INIS)

    Nester, K.

    1975-01-01

    The computer program for the calculation of the rise of cooling tower plumes (3-dimensional) was extented. In addition to the distributions of the vertical velocity, the temperatures and the specific humidity, it yields now the distribution of the rain droplets in the plume, too. The treatment of the cloud physics was based on the theory of Kessler (E. Kessler, Meteorological Monographs, 10 (1969) No. 32). (orig.) [de

  6. Improvements in or relating to cooling systems for nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Ljubivy, A.G.; Batjukov, V.I.; Shkhian, T.G.; Fadeev, A.I.

    1980-01-01

    A cooling system is proposed which can be used to cool a set of nuclear fuel assemblies arranged in a reactor core or placed in a container for spent fuel assemblies. The object of the invention is to provide a system which would prevent leakage of coolant from the vessel in the event of a rupture of the coolant supply pipeline externally of the vessel. In the case of the reactor cooling system the level of the coolant is stopped from dropping below the level of the active portion of the fuel assemblies and thus prevents a breakdown of the reactor. (UK)

  7. Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nagy, G. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary)]. E-mail: nagyg@sunserv.kfki.hu; Somogyi, A. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary); Patek, G. [Paks Nuclear Power Plant, P.O. Box 71, Paks H-7031 (Hungary); Pinter, T. [Paks Nuclear Power Plant, P.O. Box 71, Paks H-7031 (Hungary); Schiller, R. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary)

    2007-06-15

    Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.

  8. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  9. Loss of cooling accident simulation of nuclear power station spent-fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, M.; Liang, K-S., E-mail: mlee@ess.nthu.edu.tw, E-mail: ksliang_1@hotmail.com [National Tsing Hua Univ., Hsinchu, Taiwan (China); Lin, K-Y., E-mail: syrup760914@gmail.com [Taiwan Power Company, Taiwan (China)

    2014-07-01

    The core melt down accident of Fukushima Nuclear Power Station on March 11th, 2011 alerted nuclear industry that the long term loss of cooling of spent fuel pool may need some attention. The target plant analyzed is the Chinshan Nuclear Power Station of Taiwan Power Company. The 3-Dimensional RELAP5 input deck of the spent fuel pool of the station is built. The results indicate that spent fuel of Chinshan Nuclear Power Station is uncovered at 6.75 days after an accident of loss cooling takes place and cladding temperature rises above 2,200{sup o}F around 8 days. The time is about 13 hours earlier than the results predicted using simple energy balance method. The results also show that the impact of Counter Current Flow Limitation (CCFL) and radiation heat transfer model is marginal. (author)

  10. Monte Carlo Simulation of Adiabatic Cooling and Nuclear Magnetism

    DEFF Research Database (Denmark)

    Lindgård, Per-Anker; Viertiö, H. E.; Mouritsen, Ole G.

    1988-01-01

    in experimental studies of nuclear magnetism using adiabatic demagnetization methods. It is found that, although fluctuations reduce the transition temperatures by 40%, the isentropes are reduced by less than 10% relative to those calculated by mean-field theory. The dynamics of the ordering process following...

  11. A novel nuclear combined power and cooling system integrating high temperature gas-cooled reactor with ammonia–water cycle

    International Nuclear Information System (INIS)

    Luo, Chending; Zhao, Fuqiang; Zhang, Na

    2014-01-01

    Highlights: • We propose a novel nuclear ammonia–water power and cooling cogeneration system. • The high temperature reactor is inherently safe, with exhaust heat fully recovered. • The thermal performances are improved compared with nuclear combined cycle. • The base case attains an energy efficiency of 69.9% and exergy efficiency of 72.5%. • Energy conservation and emission reduction are achieved in this cogeneration way. - Abstract: A nuclear ammonia–water power and refrigeration cogeneration system (NAPR) has been proposed and analyzed in this paper. It consists of a closed high temperature gas-cooled reactor (HTGR) topping Brayton cycle and a modified ammonia water power/refrigeration combined bottoming cycle (APR). The HTGR is an inherently safe reactor, and thus could be stable, flexible and suitable for various energy supply situation, and its exhaust heat is fully recovered by the mixture of ammonia and water in the bottoming cycle. To reduce exergy losses and enhance outputs, the ammonia concentrations of the bottoming cycle working fluid are optimized in both power and refrigeration processes. With the HTGR of 200 MW thermal capacity and 900 °C/70 bar reactor-core-outlet helium, the system achieves 88.8 MW net electrical output and 9.27 MW refrigeration capacity, and also attains an energy efficiency of 69.9% and exergy efficiency of 72.5%, which are higher by 5.3%-points and 2.6%-points as compared with the nuclear combined cycle (NCC, like a conventional gas/steam power-only combined cycle while the topping cycle is a closed HTGR Brayton cycle) with the same nuclear energy input. Compared with conventional separate power and refrigeration generation systems, the fossil fuel saving (based on CH 4 ) and CO 2 emission reduction of base-case NAPR could reach ∼9.66 × 10 4 t/y and ∼26.6 × 10 4 t/y, respectively. The system integration accomplishes the safe and high-efficiency utilization of nuclear energy by power and refrigeration

  12. Electrical supply and controls for induced-draft cooling towers at Browns Ferry Nuclear Plant

    International Nuclear Information System (INIS)

    Mock, C.H.; Boehms, J.H.

    1975-01-01

    Design considerations are given for selection of electrical features as required for addition of mechanical-draft-type cooling towers at an existing multiunit nuclear generating station. Environmental and nuclear safety problems were solved economically by use of enclosed 161-kV power connections, oil-filled transformers, supervisory-type control, and unique schemes for redundancy to minimize need for Class 1E construction

  13. In-Service Inspection Approaches for Lead-Cooled Nuclear Reactors

    Science.gov (United States)

    2017-06-01

    heavily regulated and mature. For example, the Illinois Emergency Management Agency (IEMA) conducted 805 soil samples testing for radionuclides around... radiation , and lead-cooled reactors are expected to have economic advantages compared to other nuclear coolant/moderator systems due to design...their six nuclear reactors in 22 2015 (IEMA, 2016, 3). In addition, they currently have 1649 environmental dosimeters testing for gamma radiation

  14. A combined gas cooled nuclear reactor and fuel cell cycle

    Science.gov (United States)

    Palmer, David J.

    Rising oil costs, global warming, national security concerns, economic concerns and escalating energy demands are forcing the engineering communities to explore methods to address these concerns. It is the intention of this thesis to offer a proposal for a novel design of a combined cycle, an advanced nuclear helium reactor/solid oxide fuel cell (SOFC) plant that will help to mitigate some of the above concerns. Moreover, the adoption of this proposal may help to reinvigorate the Nuclear Power industry while providing a practical method to foster the development of a hydrogen economy. Specifically, this thesis concentrates on the importance of the U.S. Nuclear Navy adopting this novel design for its nuclear electric vessels of the future with discussion on efficiency and thermodynamic performance characteristics related to the combined cycle. Thus, the goals and objectives are to develop an innovative combined cycle that provides a solution to the stated concerns and show that it provides superior performance. In order to show performance, it is necessary to develop a rigorous thermodynamic model and computer program to analyze the SOFC in relation with the overall cycle. A large increase in efficiency over the conventional pressurized water reactor cycle is realized. Both sides of the cycle achieve higher efficiencies at partial loads which is extremely important as most naval vessels operate at partial loads as well as the fact that traditional gas turbines operating alone have poor performance at reduced speeds. Furthermore, each side of the cycle provides important benefits to the other side. The high temperature exhaust from the overall exothermic reaction of the fuel cell provides heat for the reheater allowing for an overall increase in power on the nuclear side of the cycle. Likewise, the high temperature helium exiting the nuclear reactor provides a controllable method to stabilize the fuel cell at an optimal temperature band even during transients helping

  15. Cooling tower make-up water processing for nuclear power plants: a comparison

    Energy Technology Data Exchange (ETDEWEB)

    Andres, O; Flunkert, F; Hampel, G; Schiffers, A [Rheinisch-Westfaelisches Elektrizitaetswerk A.G., Essen (Germany, F.R.)

    1977-01-01

    In water-cooled nuclear power plants, 1 to 2% of the total investment costs go to cooling tower make-up water processing. The crude water taken from rivers or stationary waters for cooling must be sufficiently purified regarding its content of solids, carbonate hardness and corrosive components so as to guarantee an operation free of disturbances. At the same time, the processing methods must be selected for operational-economic reasons in such a manner that waste water and waste problems are kept small regarding environmental protection. The various parameters described have a decisive influence on the processing methods of the crude water, individual processes (filtration, sedimentation, decarbonization) are described, circuit possibilities for cooling water systems are compared and the various processes are analyzed and compared with regard to profitableness and environmental compatability.

  16. Minimization of radioactive material deposition in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Ruiz, C.P.; Blaies, D.M.

    1988-01-01

    This patent describes the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering the water-bearing vessel. The improvement contains a substantially lower proportion of the /sup 64/Zn isotope than naturally occurring zinc

  17. Induction apparatus monitoring structural strains in liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Dean, S.A.; Evans, R.A.

    1981-01-01

    An improved method of monitoring induced torsional and linear strains in the internal structures of liquid metal cooled nuclear reactors is described. An electrical induction apparatus indicates the variation of magnetic coupling caused by a ferromagnetic member of the apparatus being subjected to such strains. (U.K.)

  18. Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-01-01

    Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

  19. Dry well cooling systems in BWR type nuclear power plants

    International Nuclear Information System (INIS)

    Hanamura, Ikuo; Tada, Kenji.

    1986-01-01

    Purpose: To prevent the damages of pipeways due to salt damages at the surface of control rod drives in BWR type reactors. Constitution: In control rod drives and the lowermost area in the dry well in which surface corrosion and pitching have been resulted by the salt contents in air due to the increase in the humidity accompanying the lowering of the temperature, a blower is disposed to the upstream of the cooling coils and a portion of high temperature air returned to the lower cooler is replaced with a low temperature feed air to increase the feed temperature in the area. Further, by upwardly turning the downwarded feed air drawing port in which cold feed air has so far been descended as it is, the descendance of the cold air is suppressed. As a result, temperature lowering in the driving mechanisms and the lower area can be prevented to obtain a predetermined temperature, whereby the dewing on the surface can be prevented and thereby preventing the occurrence of corrosion and pitching. (Horiuchi, T.)

  20. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    Chesworth, G.; Hind, J.R.; Hodgson, D.; Seed, G.

    1981-01-01

    In a nuclear reactor of the pool kind the primary vessel and fuel assembly are carried from the roof of the containment vault by tie straps. The primary vessel incorporates an annular yoke of 'k' cross-section the tie straps being attached to the upwardly directed vertical leg and the downwardly directed inclined leg. The upper and lower strakes of the primary vessel are extensions of the remaining legs. Load supporting welds therefore are of intermittent nature thereby limiting the effects of weld crack propagation

  1. Environmental effects of large discharges of cooling water. Experiences from Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Ehlin, Ulf; Lindahl, Sture; Neuman, Erik; Sandstroem, Olof; Svensson, Jonny

    2009-07-01

    Monitoring the environmental effects of cooling water intake and discharge from Swedish nuclear power stations started at the beginning of the 1960s and continues to this day. In parallel with long-term monitoring, research has provided new knowledge and methods to optimise possible discharge locations and design, and given the ability to forecast their environmental effects. Investigations into the environmental effects of cooling-water are a prerequisite for the issuing of power station operating permits by the environmental authorities. Research projects have been carried out by scientists at universities, while the Swedish Environmental Protection Agency, the Swedish Board of Fisheries, and the Swedish Meteorological and Hydrological Institute, SMHI, are responsible for the greater part of the investigations as well as of the research work. The four nuclear power plants dealt with in this report are Oskarshamn, Ringhals, Barsebaeck and Forsmark. They were taken into operation in 1972, 1975, 1975 and 1980 resp. - a total of 12 reactors. After the closure of the Barsebaeck plants in 2005, ten reactors remain in service. The maximum cooling water discharge from the respective stations was 115, 165, 50 and 135 m 3 /s, which is comparable to the mean flow of an average Swedish river - c:a 150 m 3 /s. The report summarizes studies into the consequences of cooling water intake and discharge. Radiological investigations made at the plants are not covered by this review. The strategy for the investigations was elaborated already at the beginning of the 1960s. The investigations were divided into pre-studies, baseline investigations and monitoring of effects. Pre-studies were partly to gather information for the technical planning and design of cooling water intake and outlet constructions, and partly to survey the hydrographic and ecological situation in the area. Baseline investigations were to carefully map the hydrography and ecology in the area and their natural

  2. IAEA activities in technology development for advanced water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

    2003-01-01

    As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

  3. Nuclear powerplant with closed gas-cooling circuit

    International Nuclear Information System (INIS)

    Haferkamp, D.; Hodzic, A.; Winter, U.

    1976-01-01

    Disclosed is a nuclear power plant comprising a pressure-tight safety vessel surrounding the entire plant, an inner vessel of reinforced concrete, a high-temperature reactor contained in the inner vessel, a gas turbine assembly having a turbine and a high- and low-pressure compressor located in a horizontally oriented chamber below the reactor, a plurality of heat exchange units positioned in a plurality of vertically oriented pods spaced radially symmetrically about the reactor and suitable conduits for carrying the reactive gas between the system components. The conduits are arranged in generally horizontally and vertically oriented straight lines, and the conduits for carrying low-pressure gas comprise a horizontal system positioned beneath the turbine assembly having a plurality of coaxial connecting tubes, collectors and distributors as well as normal conduits, so that high pressure gas flows in the internal passage and low-pressure gas flows in the outer passage. 22 claims, 7 figures

  4. The discussion of nuclear power plant's cooling chain design for freezing site

    International Nuclear Information System (INIS)

    Hu Jian; Yang Ting; Jiang Xulun

    2014-01-01

    The Component cooling water system (RRI) and Essential service water system (SEC) are composed of Nuclear Power Plant's (NPP) cooling chain, which has its special requirement for freezing site from system design and safety point of view. The feature and difficulty of cooling chain design at freezing condition (when the intake water temperature is below O ℃) are represented. At present, several NPPs are in operation or under construction at freezing site in the world, including Pressurized Water Reactor (PWR) and Canadian Deuterium Uranium reactor (CANDU). By analyzing the thoughts and applicability of different kinds of cooling chain design at freezing site, one solution called 'SEC thermal discharge reflux' is proposed to remove the residual heat from Nuclear Island (NI) into heat sink safely in winter. The solution has been approved by National Nuclear Safety Administration (NNSA) in China and applied in one of CPR NPP in the north of China, which is able to solve several problems compared with the traditional solutions, such as 'Reactor low power operation', 'Reactor start-up for the first time', and 'Changeover of RRI/SEC trains in winter'. The solution is also able to prevent RRI/SEC heat exchanger from icing and avoid low flowrate in SEC pipes. Besides, considering of the economical efficiency, simple operation and control strategy is designed. (authors)

  5. Fundamental design bases for independent core cooling in Swedish nuclear power reactors

    International Nuclear Information System (INIS)

    Jelinek, Tomas

    2015-01-01

    New regulations on design and construction of nuclear power plants came into force in 2005. The need of an independent core cooling system and if the regulations should include such a requirement was discussed. The Swedish Radiation Safety authority (SSM) decided to not include such a requirement because of open questions about the water balance and started to investigate the consequences of an independent core cooling system. The investigation is now finished and SSM is also looking at the lessons learned from the accident in Fukushima 2011. One of the most important measures in the Swedish national action plan is the implementation of an independent core cooling function for all Swedish power plants. SSM has investigated the basic design criteria for such a function where some important questions are the level of defence in depth and the acceptance criteria. There is also a question about independence between the levels of defence in depth that SSM have included in the criteria. Another issue that has to be taken into account is the complexity of the system and the need of automation where independence and simplicity are very strong criteria. In the beginning of 2014 a memorandum was finalized regarding fundamental design bases for independent core cooling in Swedish nuclear power reactors. A decision based on this memorandum with an implementation plan will be made in the first half of 2014. Sweden is also investigating the possibility to have armed personnel on site, which is not allowed currently. The result from the investigation will have impact on the possibility to use mobile equipment and the level of protection of permanent equipment. In this paper, SSM will present the memorandum for design bases for independent core cooling in Swedish nuclear power reactors that was finalized in March 20147 that also describe SSM's position regarding independence and automation of the independent core cooling function. This memorandum describes the Swedish

  6. Linear programming optimization of nuclear energy strategy with sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Lee, Je Whan; Jeong, Yong Hoon; Chang, Yoon Il; Chang, Soon Heung

    2011-01-01

    Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters

  7. Material for shutting down gas cooled nuclear reactors

    International Nuclear Information System (INIS)

    Jackson, F.

    1977-01-01

    Some disadvantage of conventional emergency shutdown means for nuclear reactors employing a supply of B steel shot or B powder are mentioned. With regard to B powder it is stated that there is some uncertainty as to whether the powder once dispersed into the core will settle in the active part of the core in sufficient quantities to ensure shutdown. The system described aims to avoid these disadvantages. Pellets are provided comprising a solid neutron poison material and a solid organic substance that remains solid at the relatively low temperature normally expected to prevail in the reactor coolant channel away from the reactor core. The organic substance melts at a higher temperature expected to prevail in the coolant channel within the core., and is adherent on melting to the coolant channel wall and to the solid neutron poison, being thus capable of causing adherence of the latter to the coolant channel wall in the reactor core. The pellets are preferably given a moisture resistant coating to prevent them sticking together and to impart free flowing characteristics. The neutron poison may consist of B, Cd, Gd, or their compounds, and for the coating a suitable polymer may be used. Steel filings may be incorporated in the pellets to aid easy flowing under gravity. Examples of manufacture of the pellets are given. (U.K.)

  8. Study on nuclear analysis method for high temperature gas-cooled reactor and its nuclear design (Thesis)

    International Nuclear Information System (INIS)

    Goto, Minoru

    2015-03-01

    An appropriate configuration of fuel and reactivity control equipment in a nuclear reactor core, which allows the design of the nuclear reactor core for low cost and high performance, is performed by nuclear design with high accuracy. The accuracy of nuclear design depends on a nuclear data library and a nuclear analysis method. Additionally, it is one of the most important issues for the nuclear design of a High Temperature Gas-cooled Reactor (HTGR) that an insertion depth of control rods into the reactor core should be retained shallow by reducing excess reactivity with a different method to keep fuel temperature below its limitation thorough a burn-up period. In this study, using experimental data of the High Temperature engineering Test Reactor (HTTR), which is a Japan's HTGR with 30 MW of thermal power, the following issues were investigated: applicability of nuclear data libraries to nuclear analysis for HTGRs; applicability of the improved nuclear analysis method for HTGRs; and effectiveness of a rod-type burnable poison on HTGR reactivity control. A nuclear design of a small-sized HTGR with 50 MW of thermal power (HTR50S) was performed using these results. In the nuclear design of the HTR50S, we challenged to decrease the kinds of the fuel enrichments and to increase the power density compared with the HTTR. As a result, the nuclear design was completed successfully by reducing the kinds of the fuel enrichment to only three from twelve of the HTTR and increasing the power density by 1.4 times as much as that of the HTTR. (author)

  9. Design and analysis of new prestressed concrete containment and its passive cooling system for nuclear power plants

    International Nuclear Information System (INIS)

    Tan Xiaoshi; Li Xiaowei; Li Xiaotian; He Shuyan

    2014-01-01

    A new nuclear power plant prestressed concrete containment and its passive cooling system design were proposed for CAP1700 nuclear power plant as an example. The thermal-hydraulic calculation method for the new passive containment cooling system of CAP1700 was introduced and the operating parameters in accident condition were obtained. The result shows that the design of passive containment cooling system for CAP1700 is feasible and can meet the cooling demand in accident condition. Reservoir capacity of tank has a big margin and can be further optimized by calculation. (authors)

  10. THE COOLING OF THE CASSIOPEIA A NEUTRON STAR AS A PROBE OF THE NUCLEAR SYMMETRY ENERGY AND NUCLEAR PASTA

    Energy Technology Data Exchange (ETDEWEB)

    Newton, William G.; Hooker, Joshua; Li, Bao-An [Department of Physics and Astronomy, Texas A and M University-Commerce, Commerce, TX 75429-3011 (United States); Murphy, Kyleah [Umpqua Community College, Roseburg, OR 97470 (United States)

    2013-12-10

    X-ray observations of the neutron star (NS) in the Cas A supernova remnant over the past decade suggest the star is undergoing a rapid drop in surface temperature of ≈2%-5.5%. One explanation suggests the rapid cooling is triggered by the onset of neutron superfluidity in the core of the star, causing enhanced neutrino emission from neutron Cooper pair breaking and formation (PBF). Using consistent NS crust and core equations of state (EOSs) and compositions, we explore the sensitivity of this interpretation to the density dependence of the symmetry energy L of the EOS used, and to the presence of enhanced neutrino cooling in the bubble phases of crustal ''nuclear pasta''. Modeling cooling over a conservative range of NS masses and envelope compositions, we find L ≲ 70 MeV, competitive with terrestrial experimental constraints and other astrophysical observations. For masses near the most likely mass of M ≳ 1.65 M {sub ☉}, the constraint becomes more restrictive 35 ≲ L ≲ 55 MeV. The inclusion of the bubble cooling processes decreases the cooling rate of the star during the PBF phase, matching the observed rate only when L ≲ 45 MeV, taking all masses into consideration, corresponding to NS radii ≲ 11 km.

  11. Study on extreme high temperature of cooling water in Chinese coastal nuclear power plant

    International Nuclear Information System (INIS)

    Yu Fan; Jiang Ziying

    2012-01-01

    In order to protect aquatic life from the harmful effects of thermal discharge, the appropriate water temperature limits or the scope of the mixing zone is a key issue in the regulatory control of the environmental impact of thermal discharge. Based on the sea surface temperature in the Chinese coastal waters, the extreme value of the seawater temperature change was analyzed by using the Gumbel model. The limit of the design temperature rise of cooling water in the outfall is 9 ℃, and the limit of the temperature rise of cooling water in the edge of the mixing zone is 4 ℃. The extreme high temperature of the cooling water in Chinese coastal nuclear power plant is 37 ℃ in the Bohai Sea, Yellow Sea, and is 40 ℃ in East China Sea, South China Sea. (authors)

  12. Method and plant to remote tritium from the cooling water of a nuclear reactor

    International Nuclear Information System (INIS)

    O'Brien, C.J.

    1976-01-01

    A method is proposed for the extraction of tritium from the cooling water of a nuclear reactor, based on the principle of concentrating the tritium by a multi-stage transfer process. The cooling water is brought into contact in each stage with basic, labile, hydrogen-containing material with high pH value, whereby the tritium is transfered into an intermediate solid product and can be separated off. The technical details of the plant are described. Cellulose materials, such as cotton and wood as well as protein-containing material, such as muscle tissue are mentioned as examples of materials with a high affinity to tritium, greater than the affinity of water to tritium. They extract tritium from the cooling water. (HK) [de

  13. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-01-01

    The age of nuclear waste - the length of time between its removal from the reactor cores and its emplacement in a repository - is a significant factor in determining the thermal loading of a repository. The surface cooling period as well as the density and sequence of waste emplacement affects both the near-field repository structure and the far-field geologic environment. To investigate these issues, a comprehensive review was made of the available literature pertaining to thermal effects and thermal properties of mined geologic repositories. This included a careful evaluation of the effects of different surface cooling periods of the wastes, which is important for understanding the optimal thermal loading of a repository. The results led to a clearer understanding of the importance of surface cooling in evaluating the overall thermal effects of a radioactive waste repository. The principal findings from these investigations are summarized in this paper

  14. Building concept of cooling towers for WWER-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Bucha, V.; David, M.

    1984-01-01

    A project is described of cooling towers with natural draught for the Temelin nuclear power plant. The concept proceeds from the classical design of the so-called Itterson type, i.e., the outer cladding of the draught stack is made of a monolithic reinforced concrete unit in the shape of a hyperboloid of revolution supported by a system of oblique supports mounted along the edge of the cooled water tank. The procedure is explained of the thermal calculation for the given operating conditions. The basic alternatives are considered of the choice of material and design of the cooling system. Questions are discussed relating to the design of the eliminator, the windwart wall and the shape of the shell of the draught stack and its loading by wind and seismic effects. (E.S.)

  15. Ecology of Legionella within water cooling circuits of nuclear power plants along the French Loire River

    International Nuclear Information System (INIS)

    Jakubek, Delphine

    2012-01-01

    The cooling circuits of nuclear power plants, by their mode of operating, can select thermophilic microorganisms including the pathogenic organism Legionella pneumophila. To control the development of this genus, a disinfection treatment of water cooling systems with monochloramine can be used. To participate in the management of health and environmental risks associated with the physico-chemical and microbiological modification of water collected from the river, EDF is committed to a process of increasing knowledge about the ecology of Legionella in cooling circuits and its links with its environment (physical, chemical and microbiological) supporting or not their proliferation. Thus, diversity and dynamics of culturable Legionella pneumophila were determined in the four nuclear power plants along the Loire for a year and their links with physico-chemical and microbiological parameters were studied. This study revealed a high diversity of Legionella pneumophila subpopulations and their dynamic seems to be related to the evolution of a small number of subpopulations. Legionella subpopulations seem to maintain strain-specific relationships with biotic parameters and present different sensitivities to physico-chemical variations. The design of cooling circuits could impact the Legionella community. The use of monochloramine severely disrupts the ecosystem but does not select biocide tolerant subpopulations. (author)

  16. Method of inhibiting concentration of radioactive corrosion products in cooling water or nuclear power plants

    International Nuclear Information System (INIS)

    Takabayashi, Jun-ichi; Hishida, Mamoru; Ishikura, Takeshi.

    1979-01-01

    Purpose: To suppress the increase in the concentration of the radioactive corrosion products in cooling water, which increase is accompanied by the transference of the corrosion products activated and accumulated in the core due to dissolution and exfoliation into the core water, and inhibit the flowing of said products out of the core and the diffusion thereof into the cooling system, thereby to prevent the accumulation of said products in the cooling system and prevent radioactive contaminations. Method: In a nuclear power plant of a BWR type light water reactor, when the temperature of the pile water is t 0 C, hydrogen is injected in cooling water in a period of time from immediately before starting of the drive stopping operation of the nuclear power plant to immediately after the termination of restarting operation, whereby the concentration of hydrogen in the reactor water through said period is maintained at a value more than 2exp (0.013 t) cm 3 N.T.P./kg H 2 O. (Aizawa, K.)

  17. Strainer device for an emergency cooling system in a nuclear power plant

    International Nuclear Information System (INIS)

    Trybom, J.

    1997-01-01

    The invention relates to a strainer device for separating contaminants from water in an emergency cooling system for a nuclear power plant. The nuclear power plant has a wet-well for water in the emergency cooling system and the strainer device comprises at least one strainer device, which is arranged in the wet-well. According to the invention the strainer is suspended in a desired position in the wet-well by means of at least a group of at least three tie rods arranged at angles to each other, each tie rod being fixed at one end to the strainer and its other end to the container or an anchor ring joined thereto. (author) figs

  18. Plugging inaccessible leaks in cooling water pipework in nuclear power plants

    International Nuclear Information System (INIS)

    Powell, A.B.; May, R.; Down, M.G.

    1988-01-01

    The manifestation of initially small leaks in ancilliary reactor cooling water systems is not an unusual event. Often these leaks are in virtually inaccessible locations - for example, buried in thick concrete shielding or situated in cramped and highly radioactive vaults. Such leaks may ultimately prejudice the availability of the entire nuclear system. Continued operation without repair can result in the leak becoming larger, and the leaking water can cause further corrosion problems and interfere with instrumentation. In addition, the water may increase the volume of radwaste. In short, initially trivial leaks may cause significant operating problems. This paper describes the sealing of such leaks in the biological shield cooling system of Ontario Hydro's Pickering nuclear generating station CANDU reactors

  19. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  20. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  1. Steam generator tube failures: world experience in water-cooled nuclear power reactors in 1975

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-11-01

    Steam generator tube failures were reported in 22 out of 62 water-cooled nuclear power plants surveyed in 1975. This was less than in 1974, and the number of the tubes affected was noticeably less. This report summarizes these failures, most of which were due to corrosion. Secondary-water chemistry control, procedures for inspection and repair, tube materials, and failure rates are discussed. (author)

  2. Nuclear reactor equipped with a flooding tank and a residual heat removal and emergency cooling system

    International Nuclear Information System (INIS)

    Schabert, H.P.; Winkler, F.

    1975-01-01

    A description is given of a nuclear reactor such as a pressurized-water reactor or the like which is equipped with a flooding tank and a residual heat removal and emergency cooling system. The flooding tank is arranged within the containment shell at an elevation above the upper edge of the reactor core and contains a liquid for flooding the reactor core in the event of a loss of coolant

  3. Thermionic nuclear reactor with internal heat distribution and multiple duct cooling

    Science.gov (United States)

    Fisher, C.R.; Perry, L.W. Jr.

    1975-11-01

    A Thermionic Nuclear Reactor is described having multiple ribbon-like coolant ducts passing through the core, intertwined among the thermionic fuel elements to provide independent cooling paths. Heat pipes are disposed in the core between and adjacent to the thermionic fuel elements and the ribbon ducting, for the purpose of more uniformly distributing the heat of fission among the thermionic fuel elements and the ducts.

  4. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

    Pathania, R.S.; Tatone, O.S.

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  5. Emergency Cooling of Nuclear Power Plant Reactors With Heat Removal By a Forced-Draft Cooling Tower

    Energy Technology Data Exchange (ETDEWEB)

    Murav’ev, V. P., E-mail: murval1@mail.ru

    2016-07-15

    The feasibility of heat removal during emergency cooling of a reactor by a forced-draft cooling tower with accumulation of the peak heat release in a volume of precooled water is evaluated. The advantages of a cooling tower over a spray cooling pond are demonstrated: it requires less space, consumes less material, employs shorter lines in the heat removal system, and provides considerably better protection of the environment from wetting by entrained moisture.

  6. Environmental Problems Associated With Decommissioning The Chernobyl Nuclear Power Plant Cooling Pond

    International Nuclear Information System (INIS)

    Farfan, E. B.; Jannik, G. T.; Marra, J. C.; Oskolkov, B. Ya.; Bondarkov, M. D.; Gaschak, S. P.; Maksymenko, A. M.; Maksymenko, V. M.; Martynenko, V. I.

    2009-01-01

    Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning

  7. The effects of age on nuclear power plant containment cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A. [Brookhaven National Lab., Upton, NY (United States); Davis, J. [Science Applications International Corp., New York, NY (United States)

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated.

  8. Environmental Problems Associated With Decommissioning The Chernobyl Nuclear Power Plant Cooling Pond

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, E. B.; Jannik, G. T.; Marra, J. C.; Oskolkov, B. Ya.; Bondarkov, M. D.; Gaschak, S. P.; Maksymenko, A. M.; Maksymenko, V. M.; Martynenko, V. I.

    2009-11-09

    Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

  9. ENVIRONMENTAL PROBLEMS ASSOCIATED WITH DECOMMISSIONING THE CHERNOBYL NUCLEAR POWER PLANT COOLING POND

    Energy Technology Data Exchange (ETDEWEB)

    Farfan, E.

    2009-09-30

    Decommissioning of nuclear power plants and other nuclear fuel cycle facilities has been an imperative issue lately. There exist significant experience and generally accepted recommendations on remediation of lands with residual radioactive contamination; however, there are hardly any such recommendations on remediation of cooling ponds that, in most cases, are fairly large water reservoirs. The literature only describes remediation of minor reservoirs containing radioactive silt (a complete closure followed by preservation) or small water reservoirs resulting in reestablishing natural water flows. Problems associated with remediation of river reservoirs resulting in flooding of vast agricultural areas also have been described. In addition, the severity of environmental and economic problems related to the remedial activities is shown to exceed any potential benefits of these activities. One of the large, highly contaminated water reservoirs that require either remediation or closure is Karachay Lake near the MAYAK Production Association in the Chelyabinsk Region of Russia where liquid radioactive waste had been deep well injected for a long period of time. Backfilling of Karachay Lake is currently in progress. It should be noted that secondary environmental problems associated with its closure are considered to be of less importance since sustaining Karachay Lake would have presented a much higher radiological risk. Another well-known highly contaminated water reservoir is the Chernobyl Nuclear Power Plant (ChNPP) Cooling Pond, decommissioning of which is planned for the near future. This study summarizes the environmental problems associated with the ChNPP Cooling Pond decommissioning.

  10. The effects of age on nuclear power plant containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A.; Davis, J.

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated

  11. Nuclear reactor, its cooling facility, nuclear power plant, and method of operating the same

    International Nuclear Information System (INIS)

    Tate, Hitoshi; Tominaga, Kenji; Fujii, Tadashi.

    1993-01-01

    The upper surface of inner structural materials in a container is partitioned by concrete structural walls to form an upper space portion. A pressure relief plate is disposed on the concrete structural walls. If an accident occurs, the pressure relief plate is operated to form a circulation path for a gas to return to the upper space portion again from the upper space portion. The temperature of cooling water in a pressure suppression chamber, that is, a wet well liquid phase portion is elevated by after heat of a reactor core. Evaporated steams transfer from the wet well gas phase portion to the upper space portion passing through pipelines and are mixed with N 2 gas present in the upper space portion. The mixed gas is cooled by a container inner wall cooled by air passing through an air cooling duct, flows downward by way of the pressure relief plate and reaches the wet well gas phase portion again. Since the gases in the upper space circulate by a driving force caused by the after heat, reliability of cooling performance can be improved upon occurrence of an accident without using an active driving force. (I.N.)

  12. Design measures to facilitate implementation of safeguards at future water cooled nuclear power plants

    International Nuclear Information System (INIS)

    1999-01-01

    The report is intended to present guidelines to the State authorities, designers and prospective purchasers of future water cooled power reactors which, if taken into account, will minimize the impact of IAEA safeguards on plant operation and ensure efficient and effective acquisition of safeguards data to the mutual benefit of the Member State, the plant operator and the IAEA. These guidelines incorporate the IAEA's experience in establishing and carrying out safeguards at currently operating nuclear power plants, the ongoing development of safeguards techniques and feedback of experience from plant operators and designers on the impact of IAEA safeguards on plant operation. The following main subjects are included: The IAEA's safeguards function for current and future nuclear power plants; summary of the political and legal foundations of the IAEA's safeguards system; the technical objective of safeguards and the supply and use of required design information; safeguards approaches for nuclear power plants; design implications of experience in safeguarding nuclear power plants and guidelines for future water cooled reactors to facilitate the implementation of safeguards

  13. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  14. Impact of nuclear data on sodium-cooled fast reactor calculations

    International Nuclear Information System (INIS)

    Aures, A.; Bostelmann, F.; Zwermann, W.; Velkov, K.

    2016-01-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors. (authors)

  15. Retrieval effects on ventilation and cooling requirements for a nuclear waste repository

    International Nuclear Information System (INIS)

    Hambley, D.F.

    1985-01-01

    The Nuclear Waste Policy Act of 1982 (Public Law 97-425) and the regulations promulgated in Title 10, Part 60 of the Code of Federal Regulations (10CFR60) by the US Nuclear Regulatory Commission (NRC) for an underground repository for spent fuel and high level nuclear waste (HLW) require that it is possible to retrieve waste, for whatever reason, from such a facility for a period of 50 years from initial storage or until the completion of the performance confirmation period, whichever comes first. This paper considers the effects that the retrievability option mandates on ventilation and cooling systems required for normal repository operations. An example is given for a hypothetical repository in salt. 18 refs., 1 tab

  16. Nuclear spin cooling by electric dipole spin resonance and coherent population trapping

    Science.gov (United States)

    Li, Ai-Xian; Duan, Su-Qing; Zhang, Wei

    2017-09-01

    Nuclear spin fluctuation suppression is a key issue in preserving electron coherence for quantum information/computation. We propose an efficient way of nuclear spin cooling in semiconductor quantum dots (QDs) by the coherent population trapping (CPT) and the electric dipole spin resonance (EDSR) induced by optical fields and ac electric fields. The EDSR can enhance the spin flip-flop rate and may bring out bistability under certain conditions. By tuning the optical fields, we can avoid the EDSR induced bistability and obtain highly polarized nuclear spin state, which results in long electron coherence time. With the help of CPT and EDSR, an enhancement of 1500 times of the electron coherence time can been obtained after a 500 ns preparation time.

  17. Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

    International Nuclear Information System (INIS)

    Liu Yu; Dong Yujie

    2011-01-01

    In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

  18. Organohalogen products from chlorination of cooling water at nuclear power stations

    International Nuclear Information System (INIS)

    Bean, R.M.

    1983-10-01

    Eight nuclear power units at seven locations in the US were studied to determine the effects of chlorine, added as a biocide, on the composition of cooling water discharge. Water, sediment and biota samples from the sites were analyzed for total organic halogen and for a variety of organohalogen compounds. Haloforms were discharged from all plants studied, at concentrations of a few μg/L (parts-per-billion). Evidence was obtained that power plants with cooling towers discharge a significant portion of the haloforms formed during chlorination to the atmosphere. A complex mixture of halogenated phenols was found in the cooling water discharges of the power units. Cooling towers can act to concentrate halogenated phenols to levels approaching those of the haloforms. Examination of samples by capillary gas chromatography/mass spectrometry did not result in identification of any significant concentrations of lipophilic base-neutral compounds that could be shown to be formed by the chlorination process. Total concentrations of lipophilic (Bioabsorbable) and volatile organohalogen material discharged ranged from about 2 to 4 μg/L. Analysis of sediment samples for organohalogen material suggests that certain chlorination products may accumulate in sediments, although no tissue bioaccumulation could be demonstrated from analysis of a limited number of samples. 58 references, 25 figures, 31 tables

  19. Activities of passive cooling applications and simulation of innovative nuclear power plant design

    International Nuclear Information System (INIS)

    Aglar, F.; Tanrykut, A.

    2002-01-01

    This paper gives a general insight on activities of the Turkish Atomic Energy Authority (TAEA) concerning passive cooling applications and simulation of innovative nuclear power plant design. The condensation mode of heat transfer plays an important role for the passive heat removal application in advanced water-cooled reactor systems. But it is well understood that the presence of noncondesable (NC) gases can greatly inhibit the condensation process due to the build up of NC gas concentration at the liquid/gas interface. The isolation condenser of passive containment cooling system of the simplified boiling water reactors is a typical application area of in-tube condensation in the presence of NC. The test matrix of the experimental investigation undertaken at the METU-CTF test facility (Middle East Technical University, Ankara) covers the range of parameters; Pn (system pressure) : 2-6 bar, Rev (vapor Reynolds number): 45,000-94,000, and Xi (air mass fraction): 0-52%. This experimental study is supplemented by a theoretical investigation concerning the effect of mixture flow rate on film turbulence and air mass diffusion concepts. Recently, TAEA participated to an international standard problem (OECD ISP-42) which covers a set of simulation of PANDA test facility (Paul Scherrer Institut-Switzerland) for six different phases including different natural circulation modes. The concept of condensation in the presence of air plays an important role for performance of heat exchangers, designed for passive containment cooling, which in turn affect the natural circulation behaviour in PANDA systems. (author)

  20. Multi-Decadal Global Cooling and Unprecedented Ozone Loss Following a Regional Nuclear Conflict

    Science.gov (United States)

    Mills, M. J.; Toon, O. B.; Lee-Taylor, J. M.; Robock, A.

    2014-12-01

    We present the first study of the global impacts of a regional nuclear war with an Earth system model including atmospheric chemistry, ocean dynamics, and interactive sea-ice and land models (Mills et al., 2014). A limited, regional nuclear war between India and Pakistan in which each side detonates 50 15-kt weapons could produce about 5 Tg of black carbon. This would self-loft to the stratosphere, where it would spread globally, producing a sudden drop in surface temperatures and intense heating of the stratosphere. Using the Community Earth System Model with the Whole Atmosphere Community Climate Model (CESM1(WACCM)), we calculate an e-folding time of 8.7 years for stratospheric black carbon, compared to 4-6.5 years for previous studies (figure panel a). Our calculations show that global ozone losses of 20-50% over populated areas, levels unprecedented in human history, would accompany the coldest average surface temperatures in the last 1000 years (figure panel c). We calculate summer enhancements in UV indices of 30-80% over Mid-Latitudes, suggesting widespread damage to human health, agriculture, and terrestrial and aquatic ecosystems. Killing frosts would reduce growing seasons by 10-40 days per year for 5 years. Surface temperatures would be reduced for more than 25 years, due to thermal inertia and albedo effects in the ocean and expanded sea ice. The combined cooling and enhanced UV would put significant pressures on global food supplies and could trigger a global nuclear famine. Knowledge of the impacts of 100 small nuclear weapons should motivate the elimination of the more than 17,000 nuclear weapons that exist today. Mills, M. J., O. B. Toon, J. Lee-Taylor, and A. Robock (2014), Multidecadal global cooling and unprecedented ozone loss following a regional nuclear conflict, Earth's Future, 2(4), 161-176, doi:10.1002/2013EF000205.

  1. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  2. Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

    Science.gov (United States)

    Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

    1993-01-01

    The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

  3. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  4. Construction and commissioning experience of evolutionary water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2004-04-01

    Electricity market liberalization is an established fact in several countries and there is a trend to adopt it in other countries. The essential aim of market liberalization is to improve the overall economic efficiency. In order that nuclear power remains a viable option for electricity generation, its costs should be competitive with alternative sources while, at the same time, it should have a safe and reliable operation record. The capital cost of nuclear power plants (NPPs) generally accounts for 43-70% of the total nuclear electricity generation costs, compared to 26-48% for coal plants and 13-32% for gas plants. Most of these expenditures are incurred during the construction phase of a NPP. The achievement of shorter construction periods using improved technology and construction methods has a significant benefit on the costs incurred prior to any production of electricity. This document is intended to make the recent worldwide experience on construction and commissioning of evolutionary water cooled NPPs available to Member States and especially to those with nuclear power plants under construction/planning, and to those seriously considering nuclear power projects in the future. The final aim is to assist utilities and other organizations in Member States to improve the construction of nuclear power plants and achieve shortened schedules and reduced costs without compromising quality and safety. This document aims to provide an overview of the most advanced technologies, methods and processes used in construction and commissioning of recent nuclear projects. To better achieve this objective the presentation is selectively focused more on the new developments rather than providing a full review of all issues related to construction and commissioning. The experience described in this TECDOC applies to managers, engineers, supervisors, technicians and workers in various organizations dealing with the site construction and commissioning of nuclear power plants

  5. 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.

  6. Method and apparatus for emergency cooling of a nuclear power plant

    International Nuclear Information System (INIS)

    Naito, Masanori; Chino, Koichi; Sato, Chikara; Inoue, Hisamichi.

    1978-01-01

    Purpose: To improve the cooling effect of spray water by eliminating the flow control effect for spray water due to increase in the steam pressure and flowing the entire spray water into the reactor core. Constitution: Upon emergency cooling of a reactor core by spraying coolants from above at the loss of coolant accident in a nuclear power plant, coolant is sprayed in a state where the temperature upon flowing into the reactor core is below the saturated temperature after heat exchange with vapors rising from the core. This enables to apply spray water always at a temperature and a flow rate in the range of whole volume falling irrespective of the water temperature in a pressure suppression pool. (Furukawa, Y.)

  7. Disordered nuclear pasta, magnetic field decay, and crust cooling in neutron stars

    Science.gov (United States)

    Horowitz, C. J.; Berry, D. K.; Briggs, C. M.; Caplan, M. E.; Cumming, A.; Schneider, A. S.

    2015-04-01

    Nuclear pasta, with non-spherical shapes, is expected near the base of the crust in neutron stars. Large scale molecular dynamics simulations of pasta show long lived topological defects that could increase electron scattering and reduce both the thermal and electrical conductivities. We model a possible low conductivity pasta layer by increasing an impurity parameter Qimp. Predictions of light curves for the low mass X-ray binary MXB 1659-29, assuming a large Qimp, find continued late time cooling that is consistent with Chandra observations. The electrical and thermal conductivities are likely related. Therefore observations of late time crust cooling can provide insight on the electrical conductivity and the possible decay of neutron star magnetic fields (assuming these are supported by currents in the crust). This research was supported in part by DOE Grants DE-FG02-87ER40365 (Indiana University) and DE-SC0008808 (NUCLEI SciDAC Collaboration).

  8. Accumulation of 137Cs in commercial fish of the Belyarsk nuclear power station cooling supply

    International Nuclear Information System (INIS)

    Trapeznikova, V.N.; Kulikov, N.V.; Trapeznikov, A.V.

    1984-01-01

    Results are presented of a comparative study of the accumulation of 137 Cs in basic species of commercial fish of the Beloyarsk reservoir which is used as the cooling supply for the Beloyarsk nuclear power station. Possible reasons for interspecies differences in accumulation of the radionuclide are indicated, and the increased accumulation of 137 Cs by free-living fish in the zone of heated water effluent from the station and the reduced accumulation of the emitter in carp, which are cultivated on artificial food in cages, are noted. Levels of the content of the radionuclide are compared in roach and farm carp from the cooling supplies of the Beloyarsk station and the Reftinsk power plant in the Urals

  9. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

  10. Natural-draught wet-type cooling tower of the Philippsburg 1 nuclear power plant

    International Nuclear Information System (INIS)

    Ernst, G.; Schnabel, G.; Bhargava, N.; Brog, P.; Caneill, J.Y.; Carhart, R.A.; Dorwarth, G.; Egler, W.; Fiedler, F.; Gassmann, F.; Haschke, D.; Hodin, A.; Hofmann, W.; Huebschmann, W.; Nester, K.; Policastro, A.J.; Rudolf, B.; Schatzmann, M.; Tinguely, M.; Vignolo, C.; Zaidineraite, M.

    1984-01-01

    In spring 1980, comprehensive field measurements were performed on the natural-draught wet type, cooling tower of Philippsburg I nuclear power plant. Performance in service and emission of cooling tower, condition of ambient atmosphere and spread of plume were studied in seven subprojects. The report on hand contains the results of the 8th subproject within which plume spreading was calculated by means of mathematical models. Efforts were made to win the participation of as large as circle of scientists as possible in order to obtain an overview on the efficiencies of the existing models. The size of visible plumes were calculated by means of emission data and ambient data and were compared with those dimensions resulting from photograph. The models and the results are described in individual reports. Results were summarized for B models. Complete data on the 16'Philippsburg incidents' are contained in the annex to the program report. (orig./HP) [de

  11. Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

    International Nuclear Information System (INIS)

    Dagard, P.; Couturier, M.

    1989-01-01

    This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

  12. Application of gas-cooled Accelerator Driven System (ADS) transmutation devices to sustainable nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A., E-mail: abanades@etsii.upm.es [ETSII/Universidad Politecnica de Madrid, J.Gutierrez Abascal, 2-28006 Madrid (Spain); Garcia, C.; Garcia, L. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba); Escriva, A.; Perez-Navarro, A. [Instituto de Ingenieria Energetica, Universidad Politecnica de Valencia, C.P. 46022 Valencia (Spain); Rosales, J. [Instituto Superior de Tecnologia y Ciencias Aplicadas. Quinta de los, Molinos, Ave. Salvador Allende y Luaces, Ciudad de la Habana, CP 10400, Apartado Postal 6163 (Cuba)

    2011-06-15

    Highlights: > Utilization of Accelerator Driven System (ADS) for Hydrogen production. > Evaluation of the potential use of gas-cooled ADS for a sustainable use of Uranium resources by transmutation of nuclear wastes, electricity and Hydrogen production. > Application of the Sulfur-Iodine thermochemical process to subcritical systems. > Application of CINDER90 to calculate burn-up in subcritical systems. - Abstract: The conceptual design of a pebble bed gas-cooled transmutation device is shown with the aim to evaluate its potential for its deployment in the context of the sustainable nuclear energy development, which considers high temperature reactors for their operation in cogeneration mode, producing electricity, heat and Hydrogen. As differential characteristics our device operates in subcritical mode, driven by a neutron source activated by an accelerator that adds clear safety advantages and fuel flexibility opening the possibility to reduce the nuclear stockpile producing energy from actual LWR irradiated fuel with an efficiency of 45-46%, either in the form of Hydrogen, electricity, or both.

  13. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  14. Static seals and their application in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    Information relative to six types of static seals commonly used in the primary cooling systems of nuclear reactors is compiled. This information includes a description of each type of seal, its material of construction, design features, operating experience, and advantages and disadvantages. The types covered include spiral-wound asbestos-filled gaskets, hollow metallic O-rings, Belleville spring type of gasketed joints, integrated elastomer and metal retainer gaskets, and solid metal gaskets with heavy cross sections. Omega, canopy, and lip seals are discussed briefly, and information on flange design for gasketing is also presented

  15. Natural circulation in water cooled nuclear power plants: Phenomena, models, and methodology for system reliability assessments

    International Nuclear Information System (INIS)

    2005-11-01

    In recent years it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. Further, the IAEA Conference on The Safety of Nuclear Power: Strategy for the Future which was convened in 1991 noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to assure that the systems perform their intended functions. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are an ongoing activity in several IAEA Member States. Some new designs also utilize natural circulation as a means to remove core power during normal operation. In response to the motivating factors discussed above, and to foster international collaboration on the enabling technology of passive systems that utilize natural circulation, an IAEA Coordinated Research Project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation was started in early 2004. Building on the shared expertise within the CRP, this publication presents extensive information on natural circulation phenomena, models, predictive tools and experiments that currently support design and analyses of natural circulation systems and highlights areas where additional research is needed. Therefore, this publication serves both to provide a description of the present state of knowledge on natural circulation in water cooled nuclear power plants and to guide the planning and conduct of the CRP in

  16. Optimization of the steam generator project of a gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Sakai, Massao

    1978-01-01

    The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

  17. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  18. Passive safety systems and natural circulation in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2009-11-01

    Nuclear power produces 15% of the world's electricity. Many countries are planning to either introduce nuclear energy or expand their nuclear generating capacity. Design organizations are incorporating both proven means and new approaches for reducing the capital costs of their advanced designs. In the future most new nuclear plants will be of evolutionary design, often pursuing economies of scale. In the longer term, innovative designs could help to promote a new era of nuclear power. Since the mid-1980s it has been recognized that the application of passive safety systems (i.e. those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially improve economics of new nuclear power plant designs. The IAEA Conference on The Safety of Nuclear Power: Strategy for the Future, which was convened in 1991, noted that for new plants 'the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate'. Some new designs also utilize natural circulation as a means to remove core power during normal operation. The use of passive systems 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. To support the development of advanced water cooled reactor designs with passive systems, investigations of natural circulation are conducted in several IAEA Member States with advanced reactor development programmes. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, the IAEA

  19. Hydrogen co-production from subcritical water-cooled nuclear power plants in Canada

    Energy Technology Data Exchange (ETDEWEB)

    Gnanapragasam, N.; Ryland, D.; Suppiah, S., E-mail: gnanapragasamn@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-06-15

    Subcritical water-cooled nuclear reactors (Sub-WCR) operate in several countries including Canada providing electricity to the civilian population. The high-temperature-steam-electrolysis process (HTSEP) is a feasible and laboratory-demonstrated large-scale hydrogen-production process. The thermal and electrical integration of the HTSEP with Sub-WCR-based nuclear-power plants (NPPs) is compared for best integration point, HTSEP operating condition and hydrogen production rate based on thermal energy efficiency. Analysis on integrated thermal efficiency suggests that the Sub-WCR NPP is ideal for hydrogen co-production with a combined efficiency of 36%. HTSEP operation analysis suggests that higher product hydrogen pressure reduces hydrogen and integrated efficiencies. The best integration point for the HTSEP with Sub-WCR NPP is upstream of the high-pressure turbine. (author)

  20. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  1. Brackish groundwater as an alternative source of cooling water for nuclear power plants in Israel

    International Nuclear Information System (INIS)

    Arad, A.; Olshina, A.

    1984-01-01

    The western Negev is being considered as a potential site for the location of a nuclear powerplant. Since this part of Israel has no surface water, the only alternatives for cooling water are piped-in water, Mediterranean water and local, brackish groundwater. The Judea Group aquifer was examined for its potential to provide the required amount of cooling water over the lifetime of the plant, without causing a drastic lowering of the regional water table. The salinity of the water tends to increase from east to west. Flow within the aquifer is in the direction of Beer Sheva, where the extraction rate is 32 to 35 million cu m/yr. This has resulted in a salinity creep of 5-10 mg Cl per year in the Beer Sheva area, which poses a danger of deterioration of its water supply in the long term. Given the assumed range of aquifer properties, extraction of brackish water for cooling purposes will not result in large changes in the regional water table. Exploitation of the more saline water to the southwest of Beer Sheva could preserve the quality of Beer Sheva's water supply, at the expense of an increase in the depth from which it must be pumped. 2 references, 7 figures, 2 tables

  2. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-11-01

    Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

  3. Technical, environmental, and socioeconomic factors associated with dry-cooled nuclear energy centers

    International Nuclear Information System (INIS)

    1976-04-01

    The report includes a review of the current state-of-the-art of dry-cooling technology for industrial and power-generating facilities and an evaluation of its technical potential and cost for large nuclear power plants. Criteria are formulated for coarse screening of the arid regions of the Western United States to select a surrogate site for more detailed site-specific analyses. The screening criteria included seismic considerations, existing transportation facilities, institutional and jurisdictional constraints, waste heat dissipation effects, water requirements, and ecologic and socioeconomic considerations. The Galt site near Las Vegas, Nevada was selected for the surrogate site analysis to assess important issues related to the construction and operation of twelve dry-cooled nuclear power plants at an arid location remote from major load centers. The assessment covers geotechnical, atmospheric and hydrologic considerations, special aspects of transporting large equipment overland to the site from seaports, analyses of potential transmission routes to major load centers, local institutional and taxing provisions, and ecologic and socioeconomic impacts

  4. Technical, environmental, and socioeconomic factors associated with dry-cooled nuclear energy centers

    Energy Technology Data Exchange (ETDEWEB)

    1976-04-01

    The report includes a review of the current state-of-the-art of dry-cooling technology for industrial and power-generating facilities and an evaluation of its technical potential and cost for large nuclear power plants. Criteria are formulated for coarse screening of the arid regions of the Western United States to select a surrogate site for more detailed site-specific analyses. The screening criteria included seismic considerations, existing transportation facilities, institutional and jurisdictional constraints, waste heat dissipation effects, water requirements, and ecologic and socioeconomic considerations. The Galt site near Las Vegas, Nevada was selected for the surrogate site analysis to assess important issues related to the construction and operation of twelve dry-cooled nuclear power plants at an arid location remote from major load centers. The assessment covers geotechnical, atmospheric and hydrologic considerations, special aspects of transporting large equipment overland to the site from seaports, analyses of potential transmission routes to major load centers, local institutional and taxing provisions, and ecologic and socioeconomic impacts.

  5. Reliability analysis of nuclear component cooling water system using semi-Markov process model

    International Nuclear Information System (INIS)

    Veeramany, Arun; Pandey, Mahesh D.

    2011-01-01

    Research highlights: → Semi-Markov process (SMP) model is used to evaluate system failure probability of the nuclear component cooling water (NCCW) system. → SMP is used because it can solve reliability block diagram with a mixture of redundant repairable and non-repairable components. → The primary objective is to demonstrate that SMP can consider Weibull failure time distribution for components while a Markov model cannot → Result: the variability in component failure time is directly proportional to the NCCW system failure probability. → The result can be utilized as an initiating event probability in probabilistic safety assessment projects. - Abstract: A reliability analysis of nuclear component cooling water (NCCW) system is carried out. Semi-Markov process model is used in the analysis because it has potential to solve a reliability block diagram with a mixture of repairable and non-repairable components. With Markov models it is only possible to assume an exponential profile for component failure times. An advantage of the proposed model is the ability to assume Weibull distribution for the failure time of components. In an attempt to reduce the number of states in the model, it is shown that usage of poly-Weibull distribution arises. The objective of the paper is to determine system failure probability under these assumptions. Monte Carlo simulation is used to validate the model result. This result can be utilized as an initiating event probability in probabilistic safety assessment projects.

  6. Valves for condenser-cooling-water circulating piping in thermal power station and nuclear power station

    International Nuclear Information System (INIS)

    Kondo, Sumio

    1977-01-01

    Sea water is mostly used as condenser cooling water in thermal and nuclear power stations in Japan. The quantity of cooling water is 6 to 7 t/sec per 100,000 kW output in nuclear power stations, and 3 to 4 t/sec in thermal power stations. The pipe diameter is 900 to 2,700 mm for the power output of 75,000 to 1,100,000 kW. The valves used are mostly butterfly valves, and the reliability, economy and maintainability must be examined sufficiently because of their important role. The construction, number and arrangement of the valves around a condenser are different according to the types of a turbine and the condenser and reverse flow washing method. Three types are illustrated. The valves for sea water are subjected to the electrochemical corrosion due to sea water, the local corrosion due to stagnant water, the fouling by marine organisms, the cavitation due to valve operation, and the erosion by earth and sand. The fundamental construction, use and features of butterfly valves are described. The cases of the failure and repair of the valves after their delivery are shown, and they are the corrosion of valve bodies and valve seats, and the separation of coating and lining. The newly developed butterfly valve with overall water-tight rubber lining is introduced. (Kako, I.)

  7. Thermal tests of large recirculation cooling installations for nuclear power plants

    Science.gov (United States)

    Balunov, B. F.; Lychakov, V. D.; Il'in, V. A.; Shcheglov, A. A.; Maslov, O. P.; Rasskazova, N. A.; Rakhimov, R. Z.; Boyarov, R. A.

    2017-11-01

    The article presents the results from thermal tests of some recirculation installations for cooling air in nuclear power plant premises, including the volume under the containment. The cooling effect in such installations is produced by pumping water through their heat-transfer tubes. Air from the cooled room is blown by a fan through a bundle of transversely finned tubes and is removed to the same room after having been cooled. The finning of tubes used in the tested installations was made of Grade 08Kh18N10T and Grade 08Kh18N10 stainless steels or Grade AD1 aluminum. Steel fins were attached to the tube over their entire length by means of high-frequency welding. Aluminum fins were extruded on a lathe from the external tube sheath into which a steel tube had preliminarily been placed. Although the fin extrusion operation was accompanied by pressing the sheath inner part to the steel tube, tight contact between them over the entire surface was not fully achieved. In view of this, the air gap's thermal resistance coefficient was introduced in calculating the heat transfer between the heat-transferring media. The air gap average thickness was determined from the test results taking into account the gap variation with temperature due to different linear expansion coefficients of steel and aluminum. These tests, which are part of the acceptance tests of the considered installations, were carried out at the NPO TsKTI test facility and were mainly aimed at checking if the obtained thermal characteristics were consistent with the values calculated according to the standard recommendations with introduction, if necessary, of modifications to those recommendations.

  8. Multi-purpose nuclear heat source for advanced gas-cooled reactor plants

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1993-01-01

    Nuclear power has the potential to be the ultimate green technology in that it could eliminate the need for burning fossil fuels with their polluting combustion products and greenhouse gases. This view is shared by many technologists, but it may be a generation before the public becomes convinced, and that will involve overcoming many safety, institutional, financial, and technical impediments. This paper addresses only the latter topic; a major theme being that for nuclear power to truly be a green technology and significantly benefit society, it must meet the needs of the full energy spectrum. Specifically, it must satisfy energy needs beyond just the electricity generating sector by today's nuclear plants. By virtue of its high temperature capability, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is the only type of reactor that has the potential to meet the wide range of energy needs that will emerge in the future. This paper discusses the nuclear heat source that gives the MHTGR multi-purpose capability, which is recognized today, but will not be implemented until early in the next century

  9. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  10. Wind tunnel experimental study on effect of inland nuclear power plant cooling tower on air flow and dispersion of pollutant

    International Nuclear Information System (INIS)

    Qiao Qingdang; Yao Rentai; Guo Zhanjie; Wang Ruiying; Fan Dan; Guo Dongping; Hou Xiaofei; Wen Yunchao

    2011-01-01

    A wind tunnel experiment for the effect of the cooling tower at Taohuajiang nuclear power plant on air flow and dispersion of pollutant was introduced in paper. Measurements of air mean flow and turbulence structure in different directions of cooling tower and other buildings were made by using an X-array hot wire probe. The effects of the cooling tower and its drift on dispersion of pollutant from the stack were investigated through tracer experiments. The results show that the effect of cooling tower on flow and dispersion obviously depends on the relative position of stack to cooling towers, especially significant for the cooling tower parallel to stack along wind direction. The variation law of normalized maximum velocity deficit and perturbations in longitudinal turbulent intensity in cooling tower wake was highly in accordance with the result of isolated mountain measured by Arya and Gadiyaram. Dispersion of pollutant in near field is significantly enhanced and plume trajectory is changed due to the cooling towers and its drift. Meanwhile, the effect of cooling tower on dispersion of pollutant depends on the height of release. (authors)

  11. The ring-stiffened shell of the ISAR II nuclear power plant natural-draught cooling tower

    International Nuclear Information System (INIS)

    Form, J.

    1986-01-01

    The natural-draught cooling tower of the ISAR II nuclear power plant is one of the largest in the world. The bid specifications provided for an unstiffened cooling tower shell. For the execution, however, it was decided to adopt a shell with three additional stiffening rings. The present contribution deals with the static and dynamic calculations of the execution and, in particular, with the working technique employed for the construction of the rings. (author)

  12. Multidecadal global cooling and unprecedented ozone loss following a regional nuclear conflict

    Science.gov (United States)

    Mills, Michael J.; Toon, Owen B.; Lee-Taylor, Julia; Robock, Alan

    2014-04-01

    We present the first study of the global impacts of a regional nuclear war with an Earth system model including atmospheric chemistry, ocean dynamics, and interactive sea ice and land components. A limited, regional nuclear war between India and Pakistan in which each side detonates 50 15 kt weapons could produce about 5 Tg of black carbon (BC). This would self-loft to the stratosphere, where it would spread globally, producing a sudden drop in surface temperatures and intense heating of the stratosphere. Using the Community Earth System Model with the Whole Atmosphere Community Climate Model, we calculate an e-folding time of 8.7 years for stratospheric BC compared to 4-6.5 years for previous studies. Our calculations show that global ozone losses of 20%-50% over populated areas, levels unprecedented in human history, would accompany the coldest average surface temperatures in the last 1000 years. We calculate summer enhancements in UV indices of 30%-80% over midlatitudes, suggesting widespread damage to human health, agriculture, and terrestrial and aquatic ecosystems. Killing frosts would reduce growing seasons by 10-40 days per year for 5 years. Surface temperatures would be reduced for more than 25 years due to thermal inertia and albedo effects in the ocean and expanded sea ice. The combined cooling and enhanced UV would put significant pressures on global food supplies and could trigger a global nuclear famine. Knowledge of the impacts of 100 small nuclear weapons should motivate the elimination of more than 17,000 nuclear weapons that exist today.

  13. Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

    2003-07-01

    Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

  14. Future needs for dry or peak shaved dry/wet cooling and significance to nuclear power plants. Final report

    International Nuclear Information System (INIS)

    Clukey, H.V.; McNelly, M.J.; Mitchell, R.C.

    1976-02-01

    U.S. requirements for uncommitted nuclear installations in water scarce areas that might require dry cooling tower systems are minimal through the year 2000 (6 to 23 GWe). In these areas it appears that peak-shaved dry/wet cooling systems are more attractive than all-dry tower cooling unless water costs were to approach the high level of several cents per gallon. The differential cooling system evaluated cost of peak-shaved dry/wet cooling systems above wet towers is typically $20 to $30/kWe for steam turbines; whereas, dry towers can represent an incremental burden of as much as $80/kWe. Gas turbine (Brayton Cycle) systems show similar benefits from an evaporative heat sink to those for steam turbine cycles--lower cooling system evaluated costs for peak-shaved dry/wet cooling systems than for conventional wet towers. These cooling system cost differentials do not reflect total costs for Brayton Cycle gas turbine plants. Together these added costs and uncertainties may substantially exceed the dollar incentives available for development of the Brayton Cycle for power generation needs for water deficient sites

  15. Computer optimization of dry and wet/dry cooling tower systems for large fossil and nuclear power plants

    International Nuclear Information System (INIS)

    Choi, M.; Glicksman, L.R.

    1979-02-01

    This study determined the cost of dry cooling compared to the conventional cooling methods. Also, the savings by using wet/dry instead of all-dry cooling were determined. A total optimization was performed for power plants with dry cooling tower systems using metal-finned-tube heat exchangers and surface condensers. The optimization minimizes the power production cost. The program optimizes the design of the heat exchanger and its air and water flow rates. In the base case study, the method of replacing lost capacity assumes the use of gas turbines. As a result of using dry cooling towers in an 800 MWe fossil plant, the incremental costs with the use of high back pressure turbine and conventional turbine over all-wet cooling are 11 and 15%, respectively. For a 1200 MWe nuclear plant, these are 22 and 25%, respectively. Since the method of making up lost capacity depends on the situation of a utility, considerable effort has been placed on testing the effects of using different methods of replacing lost capacity at high ambient temperatures by purchased energy. The results indicate that the optimization is very sensitive to the method of making up lost capacity. It is, therefore, important to do an accurate representation of all possible methods of making up capacity loss when optimizating power plants with dry cooling towers. A solution for the problem of losing generation capability by a power plant due to the use of a dry cooling tower is to supplement the dry tower during the hours of peak ambient temperatures by a wet tower. A separate wet/dry cooling tower system with series tower arrangement was considered in this study, and proved to be an economic choice over all-dry cooling where some water is available but supplies are insufficient for a totally evaporative cooling tower

  16. Fluidized bed nuclear reactor as a IV generation reactor

    International Nuclear Information System (INIS)

    Sefidvash, Farhang

    2002-01-01

    The object of this paper is to analyze the characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept under the light of the requirements set for the IV generation nuclear reactors. It is seen that FBNR generally meets the goals of providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production; minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment; increase the assurance that it is a very unattractive and least desirable route for diversion or theft of weapons-usable materials; excel in safety and reliability; have a very low likelihood and degree of reactor core damage; eliminate the need for offsite emergency response; have a clear life-cycle cost advantage over other energy sources; have a level of financial risk comparable to other energy projects. The other advantages of the proposed design are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (author)

  17. Nuclear closed-cycle gas turbine (HTGR-GT): dry cooled commercial power plant studies

    International Nuclear Information System (INIS)

    McDonald, C.F.; Boland, C.R.

    1979-11-01

    Combining the modern and proven power conversion system of the closed-cycle gas turbine (CCGT) with an advanced high-temperature gas-cooled reactor (HTGR) results in a power plant well suited to projected utility needs into the 21st century. The gas turbine HTGR (HTGR-GT) power plant benefits are consistent with national energy goals, and the high power conversion efficiency potential satisfies increasingly important resource conservation demands. Established technology bases for the HTGR-GT are outlined, together with the extensive design and development program necessary to commercialize the nuclear CCGT plant for utility service in the 1990s. This paper outlines the most recent design studies by General Atomic for a dry-cooled commercial plant of 800 to 1200 MW(e) power, based on both non-intercooled and intercooled cycles, and discusses various primary system aspects. Details are given of the reactor turbine system (RTS) and on integrating the major power conversion components in the prestressed concrete reactor vessel

  18. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    International Nuclear Information System (INIS)

    Downey, G.L.

    1988-01-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action. (author)

  19. Surveillance tests for light-water cooled nuclear power reactor vessels in IMEF

    International Nuclear Information System (INIS)

    Choo, Yong-Sun; Ahn, Sang-Bok; Park, Dae-Gyu; Jung, Yang-Hong; Yoo, Byung-Ok; Oh, Wan-Ho; Baik, Seung-Je; Koo, Dae-Seo; Lee, Key-Soon

    1999-01-01

    The surveillance tests for light-water cooled nuclear power reactor vessels were established to monitor the radiation-induced changes in the mechanical properties of ferritic materials in the beltline according to US NRC 10 CFR 50 App. G, US NRC RG1.99-rev.2, ASTM E185-82 and E185-94 in Irradiated Materials Examination Facility(IMEF). The surveillance capsule was transported from NPPs pool sites to KAERI IMEF by using a shipping cask. The capsule was cut and dismantled by capsule cutting machine and milling machine in M2 hot cell. Charpy tests and tension tests were performed in M5a and M5b hot cells respectively. Especially the EPMA located at hot lab was used to analyze the Ni and Cu wt% composition of base metal and weld for predicting the adjusted reference temperature(ART). The established process and test results were summarized in this paper. (author)

  20. KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

  1. Post-LOCA long term cooling performance in Korean standard nuclear power plants

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    1999-01-01

    The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break LOCA and large break LOCA. The RELAP5/MOD3.2.2 code is used to calculate the LTC sequences based on the LTC plan of the KSNPP. A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important manual action including the safety injection tank isolation in LTC procedure is investigated

  2. Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1981-01-01

    The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

  3. Reproducing cultural identity in negotiating nuclear power: the Union of Concerned Scientists and emergency core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Downey, G L

    1988-05-01

    This paper advances the concept of 'cultural identity' to account for the nexus between structure and practice in technological negotiations. It describes how the formation of the Union of Concerned Scientists (UCS), and that group's subsequent discourse and nonverbal actions, both reproduced the established identities of group members and contributed to negotiations that reconstituted those identities. In particular, UCS claims about emergency core-cooling systems in nuclear plants were congruent with the combination of a shared ideology, the social interests of Massachusetts Institute of Technology faculty, and established principles of engineering design. The cultural analysis of identity reproduction shows the opposition between cognitive and social phenomena to be a significant distinction framing action in Western culture. The analysis also suggests that new attention be given to the relationship between the constitutive and reproductive functions of discourse and nonverbal action.

  4. Indian experience with radionuclide transport, deposition and decontamination in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Narasimhan, S.V.; Das, P.C.; Lawrence, D.A.; Mathur, P.K.; Venkateswarlu, K.S.

    1983-01-01

    The present generation of water-cooled nuclear reactors uses construction materials chosen with utmost care so that minimum corrosion occurs during the life of the reactor. As interaction between the primary coolant and the construction materials is unavoidable, the coolant is chemically treated to achieve maximum compatibility. First measurements of the chemical and radiochemical composition of the crud present on the in-core and out-of-core primary heat transport system surfaces of a pressurized heavy-water-moderated and cooled reactor (PHWR) are given; then experience in India in the development of a low temperature, one-stage decontaminating formulation for chemical decontamination of the radioactive deposits formed on stainless steel surfaces under BWR conditions is discussed. The effect of the magnitude of the transients in parameters such as reactor power, system temperature, dissolved oxygen content in the coolant, etc. on the nature and migration behaviour of primary heat transport system crud in a PHWR is described. Contributions to radioactive sources and insoluble crud from different primary heat transport system materials are identified and correlated with reactor operations in a PHWR. Man-rem problems faced by nuclear reactors, especially during off-line maintenance, stress the need for reducing the deposited radioactive sources from system surfaces which would otherwise be accessible. Laboratory and on-site experimentation was carried out to effect chemical decontamination on the radioactive deposits formed on the stainless steel surfaces under BWR conditions. Both the reducing and oxidizing formulations were subsequently used in a small-scale, in-plant trial in the clean-up system of a BWR. More than 85% of the deposited 60 Co activity was found to have been removed by the oxidizing formulation. Efforts to develop a decontaminating mixture containing a reducing agent with the help of a circulating loop are in progress in the laboratory. (author)

  5. Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Bang, In Cheol

    2016-01-01

    Highlights: • Hybrid heat pipe was presented as a passive cooling device for dry storage cask of SNF. • A method to utilize waste heat from spent fuel was suggested using hybrid heat pipe. • CFD analysis was performed to evaluate the thermal performance of hybrid heat pipe. • Hybrid heat pipe can increase safety margin and storage capacity of the dry storage cask. - Abstract: Conventional dry storage facilities for spent nuclear fuel (SNF) were designed to remove decay heat through the natural convection of air, but this method has limited cooling capacity and a possible re-criticality accident in case of flooding. To enhance the safety and capacity of dry storage cask of SNF, hybrid heat pipe-based passive cooling device was suggested. Heat pipe is an excellent passive heat transfer device using the principles of both conduction and phase change of the working fluid. The heat pipe containing neutron absorber material, the so-called hybrid heat pipe, is expected to prevent the re-criticality accidents of SNF and to increase the safety margin during interim and long term storage period. Moreover, a hybrid heat pipe with thermoelectric module, a Stirling engine and a phase change material tank can be used for utilization of the waste heat as heat-transfer medium. Located at the guide tube or instrumentation tube, hybrid heat pipe can remove decay heat from inside the sealed metal cask to outside, decreasing fuel rod temperature. In this paper, a 2-step analysis was performed using computational fluid dynamics code to evaluate the heat and fluid flow inside a cask, which consisted of a single spent fuel assembly simulation and a full-scope dry cask simulation. For a normal dry storage cask, the maximum fuel temperature is 290.0 °C. With hybrid heat pipe cooling, the temperature decreased to 261.6 °C with application of one hybrid heat pipe per assembly, and to 195.1 °C with the application of five hybrid heat pipes per assembly. Therefore, a dry

  6. Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

    International Nuclear Information System (INIS)

    Babaev, N.S.

    1981-06-01

    The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

  7. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    International Nuclear Information System (INIS)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi

    2017-01-01

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t_8_/_5 (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  8. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi [Inner Mongolia Univ. of Science and Technology, Baotou (China). School of Material and Metallurgy; Kang, Xiaolan [Baotou Vocational and Technical College (China)

    2017-02-15

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t{sub 8/5} (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  9. Method of 16N generation for test of radiation controlled channels at nuclear power stations with water-cooled reactors

    International Nuclear Information System (INIS)

    Khryachkov, V.A.; Bondarenko, I.P.; Dvornikov, P.A.; Zhuravlev, B.V.; Kovtun, S.N.; Khromyleva, T.A.; Pavlov, A.V.; Roshchin, N.G.

    2012-01-01

    The preferences of nuclear reaction use for radiation control channels test in water-cooled power reactors have been analyzed in the paper. The new measurements for more accurate determination of reaction cross section energy dependence have been carried out. A set of new methods for background reducing and improvement of events determination reliability has also been developed [ru

  10. R and D programme on generation IV nuclear energy systems: the high temperatures gas-cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Billot, P.; Anzieu, P.; Brossard, P.

    2005-01-01

    The Generation IV Technology Roadmap selected, among others, a sequenced development of advanced high temperature gas cooled reactors as one of the main focus for R and D on future nuclear energy systems. The selection of this research objective originates both from the significance of high temperature and fast neutrons for nuclear energy to meet the needs for a sustainable development for the medium-long term (2020/2030 and beyond), and from the significant common R and D pathway that supports both medium term industrial projects and more advanced versions of gas cooled reactors. The first step of the 'Gas Technology Path' aims to support the development of a modular HTR to meet specific international market needs around 2020. The second step is a Very High Temperature Reactor - VHTR (>950 C) - to efficiently produce hydrogen through thermo-chemical or electro-chemical water splitting or to generate electricity with an efficiency above 50%, among other applications of high temperature nuclear heat. The third step of the Path is a Gas Fast Reactor - GFR - that features a fast-spectrum helium-cooled reactor and closed fuel cycle, with a direct or indirect thermodynamic cycle for electricity production and full recycle of actinides. Hydrogen production is also considered for the GFR. The paper succinctly presents the R and D program currently under definition and partially launched within the Generation IV International Forum on this consistent set of advanced gas cooled nuclear systems. (orig.)

  11. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  12. Evaluation of post-LOCA long term cooling performance in Korean standard nuclear power plants

    International Nuclear Information System (INIS)

    Bang, Young Seok; Jung, Jae Won; Seul, Kwang Won; Kim, Hho Jung

    2001-01-01

    The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from 0.02 to 0.5 ft 2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT) isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences. (author)

  13. CFD-DEM simulation of a conceptual gas-cooled fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Almeida, Lucilla C.; Su, Jian

    2015-01-01

    Several conceptual designs of the fluidized-bed nuclear reactor have been proposed due to its many advantages over conventional nuclear reactors such as PWRs and BWRs. Amongst their characteristics, the enhanced heat transfer and mixing enables a more uniform temperature distribution, reducing the risk of hot-spot and excessive fuel temperature, in addition to resulting in a higher burnup of the fuel. Furthermore, the relationship between the bed height and reactor neutronics turns the coolant flow rate control into a power production mechanism. Moreover, the possibility of removing the fuel by gravity from the movable core in case of a loss-of-cooling accident increases its safety. High-accuracy modeling of particles and coolant flow in fluidized bed reactors is needed to evaluate reliably the thermal-hydraulic efficiency and safety margin. The two-way coupling between solid and fluid can account for high-fidelity solid-solid interaction and reasonable accuracy in fluid calculation and fluid-solid interaction. In the CFD-DEM model, the particles are modeled as a discrete phase, following the DEM approach, whereas the fluid flow is treated as a continuous phase, described by the averaged Navier-Stokes equations on a computational cell scale. In this work, the coupling methodology between Fluent and Rocky is described. The numerical approach was applied to the simulation of a bubbling fluidized bed and the results were compared to experimental data and showed good agreement. (author)

  14. CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Shang Zhi

    2009-01-01

    The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles' geometry of the nuclear reactors. For satisfying the near-wall resolution of y + ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods' walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.

  15. Nuclear data uncertainty analysis for the generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Pelloni, S.; Mikityuk, K.

    2012-01-01

    For the European 2400 MW Gas-cooled Fast Reactor (GoFastR), this paper summarizes a priori uncertainties, i.e. without any integral experiment assessment, of the main neutronic parameters which were obtained on the basis of the deterministic code system ERANOS (Edition 2.2-N). JEFF-3.1 cross-sections were used in conjunction with the newest ENDF/B-VII.0 based covariance library (COMMARA-2.0) resulting from a recent cooperation of the Brookhaven and Los Alamos National Laboratories within the Advanced Fuel Cycle Initiative. The basis for the analysis is the original GoFastR concept with carbide fuel pins and silicon-carbide ceramic cladding, which was developed and proposed in the first quarter of 2009 by the 'French alternative energies and Atomic Energy Commission', CEA. The main conclusions from the current study are that nuclear data uncertainties of neutronic parameters may still be too large for this Generation IV reactor, especially concerning the multiplication factor, despite the fact that the new covariance library is quite complete; These uncertainties, in relative terms, do not show the a priori expected increase with bum-up as a result of the minor actinide and fission product build-up. Indeed, they are found almost independent of the fuel depletion, since the uncertainty associated with 238 U inelastic scattering results largely dominating. This finding clearly supports the activities of Subgroup 33 of the Working Party on International Nuclear Data Evaluation Cooperation (WPEC), i.e. Methods and issues for the combined use of integral experiments and covariance data, attempting to reduce the present unbiased uncertainties on nuclear data through adjustments based on available experimental data. (authors)

  16. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  17. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  18. Aquatic ecology of the Kadra reservoir, the source of cooling water for Kaiga nuclear power plant

    International Nuclear Information System (INIS)

    Ghosh, T.K.; Zargar, S.; Dhopte, R.; Kulkarni, A.; Kaul, S.N.

    2002-01-01

    The study is being conducted since July 2000 to evaluate impact of cooling water discharges from Kaiga Nuclear Power Plant on physicochemical and biological characteristics of Kadra reservoir. Besides marginal decrease of DO, sulfate, nitrate and potassium near discharge point at surface water, abiotic features of the water samples collected from three layers, viz. surface, 3-m depth and bottom at nine locations of the reservoir, did not show remarkable differences with reference to pH, phosphate, conductivity, suspended solids, sodium, hardness, chloride, alkalinity and heavy metals (Cu, Fe, Ni, Zn, Pb, Cd, Cr and Mn). The DT varied between 5 and 8.5 degC at surface water during the study. The abiotic characteristics of the reservoir water meet the specification of drinking water standard of Bureau of Indian Standards. While the counts of phytoplankton and zooplankton were reduced near discharge point, their population at 500 m off the discharge point was comparable to those near dam site at about 11 km down stream from plant site. Plamer's index (0-15) and Shannon's diversity index values (1.39-2.44) of the plankton at different sampling points indicate oligotrophic and semi productive nature of the water body. The total coliform (TC), staphylococcus and heterotrophic counts were, in general, less near discharge point. Based on TC count, the reservoir water, during most of the period, is categorized as 'B' following CPCB classification of surface waters. Generation of data needs to be continued till 2-3 years for statistical interpretation and drawing conclusions pertaining to extent of impact of cooling water discharges on Kadra reservoir ecology. (author)

  19. Study on the nuclear heat application system with a high temperature gas-cooled reactor and its safety evaluation (Thesis)

    International Nuclear Information System (INIS)

    Inaba, Yoshitomo

    2008-03-01

    Aiming at the realization of the nuclear heat application system with a High Temperature Gas-cooled Reactor (HTGR), research and development on the whole evaluation of the system, the connection technology between the HTGR and a chemical plant such as the safety evaluation against the fire and explosion and the control technology, and the vessel cooling system of the HTGR were carried out. In the whole evaluation of the nuclear heat application system, an ammonia production system using nuclear heat was examined, and the technical subjects caused by the connection of the chemical plant to the HTGR were distilled. After distilling the subjects, the safety evaluation method against the fire and explosion to the reactor, the mitigation technology of thermal disturbance to the reactor, and the reactor core cooling by the vessel cooling system were discussed. These subjects are very important in terms of safety. About the fire and explosion, the safety evaluation method was established by developing the process and the numerical analysis code system. About the mitigation technology of the thermal disturbance, it was demonstrated that the steam generator, which was installed at the downstream of the chemical reactor in the chemical plant, could mitigate the thermal disturbance to the reactor. In order to enhance the safety of the reactor in accidents, the heat transfer characteristic of the passive indirect core cooling system was investigated, and the heat transfer equation considering both thermal radiation and natural convection was developed for the system design. As a result, some technical subjects related to safety in the nuclear heat application system were solved. (author)

  20. Development of technique for estimating primary cooling system break diameter in predicting nuclear emergency event sequence

    International Nuclear Information System (INIS)

    Tatebe, Yasumasa; Yoshida, Yoshitaka

    2012-01-01

    If an emergency event occurs in a nuclear power plant, appropriate action is selected and taken in accordance with the plant status, which changes from time to time, in order to prevent escalation and mitigate the event consequences. It is thus important to predict the event sequence and identify the plant behavior resulting from the action taken. In predicting the event sequence during a loss-of-coolant accident (LOCA), it is necessary to estimate break diameter. The conventional method for this estimation is time-consuming, since it involves multiple sensitivity analyses to determine the break diameter that is consistent with the plant behavior. To speed up the process of predicting the nuclear emergency event sequence, a new break diameter estimation technique that is applicable to pressurized water reactors was developed in this study. This technique enables the estimation of break diameter using the plant data sent from the safety parameter display system (SPDS), with focus on the depressurization rate in the reactor cooling system (RCS) during LOCA. The results of LOCA analysis, performed by varying the break diameter using the MAAP4 and RELAP5/MOD3.2 codes, confirmed that the RCS depressurization rate could be expressed by the log linear function of break diameter, except in the case of a small leak, in which RCS depressurization is affected by the coolant charging system and the high-pressure injection system. A correlation equation for break diameter estimation was developed from this function and tested for accuracy. Testing verified that the correlation equation could estimate break diameter accurately within an error of approximately 16%, even if the leak increases gradually, changing the plant status. (author)

  1. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2011-11-01

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  2. Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Shang Zhi, E-mail: shangzhi@tsinghua.org.c [Faculty of Engineering, Kingston University, London SW15 3DW (United Kingdom); Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.co [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2010-04-15

    The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90 deg. the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  3. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  4. Improving activity transport models for water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Burrill, K.A

    2001-08-01

    Eight current models for describing radioactivity transport and radiation field growth around water-cooled nuclear power reactors have been reviewed and assessed. A frequent failing of the models is the arbitrary nature of the determination of the important processes. Nearly all modelers agree that the kinetics of deposition and release of both dissolved and particulate material must be described. Plant data must be used to guide the selection and development of suitable improved models, with a minimum of empirically-based rate constraints being used. Limiting case modelling based on experimental data is suggested as a way to simplify current models and remove their subjectivity. Improved models must consider the recent change to 'coordinated water chemistry' that appears to produce normal solubility behaviour for dissolved iron throughout the fuel cycle in PWRs, but retrograde solubility remains for dissolved nickel. Profiles are suggested for dissolved iron and nickel concentrations around the heat transport system in CANDU reactors, which operate nominally at constant chemistry, i.e., pH{sub T} constant with time, and which use carbon steel isothermal piping. These diagrams are modified for a CANDU reactor with stainless steel piping, in order to show the changes expected. The significance of these profiles for transport in PWRs is discussed for further model improvement. (author)

  5. Vibration Monitoring Using Fiber Optic Sensors in a Lead-Bismuth Eutectic Cooled Nuclear Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Ben De Pauw

    2016-04-01

    Full Text Available Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 μϵ allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

  6. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations

  7. Analysis of the Strength on the Rotor Punching Sheet of Nuclear Reactor Cooling Medium Driving Motor

    Directory of Open Access Journals (Sweden)

    GE Bao-jun

    2017-02-01

    Full Text Available A strong stress is withstood by the rotor punching sheet during the running of nuclear reactor cooling medium driving motor. In order to study the strength on the rotor punching sheet and the influential factor of its stress,the rotor of driving motor was the research object, the three-dimensional rotor model of driving motor is established by the finite element method to obtain the Mires equivalent stress nephogram and check the rotor’s strength with setting parameters and constraints. According to different rotor speeds,the different average temperatures of rotor punching sheet and shaft and the different static magnitude of interference between rotor punching sheet and shaft,the research about how the contact pressure of matching surface between rotor punching sheet and shaft and the Mires equivalent stress are impacted is carried on. The results show that the maximum Miser equivalent stress value of rotor punching sheet emerges on the axial vents,the stress value is beyond the tensile limit of the materialand. The greater the static magnitude of interference and the smaller temperature difference of rotor punching sheet and shaft lead to the greater interface compressive stress of rotor punching sheet and shaft and the greater maximum Mires equivalent stress value of rotor punching sheet. The higher the rotor speed lead to the smaller interface compressive stress of rotor punching sheet and shaft and the greater equivalent stress value of rotor punching sheet.

  8. Cooling water intake and discharge facilities for Ikata Nuclear Power Station

    International Nuclear Information System (INIS)

    Ishihara, Hisashi; Iwabe, Masakazu

    1977-01-01

    Igata Nuclear Power Station is located at the root of Sadamisaki peninsula in the western part of Ehime Prefecture, Japan, and faces the Iyonada sea area in Seto Inland Sea. The most part of the shoreline forms the cliffs, and the bottom of the sea is rather steep, reaching 60 m depth at 300 m offshore. Considering warm water discharge measures in addition to the natural conditions of tide and current, temperature of sea water, water quality and wave data, it was decided that the deep layer intake system using bottom laid intake pipes and the submerged discharge system with caisson penetrable dike would be adopted for cooling water. The latter was first employed in Japan, and the submerged discharge system with caisson penetrable dike had been developed. The intake was designed to take sea water of about 38 m 3 per sec for each condenser unit at the depth of approximately 17 m with 4.8 m diameter and 116 m length pipes and its calculation details and construction are described. The discharge system was designed to provide a horseshoe-shaped discharge pond with inner diameter of approximately 50 m, surrounded by 17 concrete caissons, and to spout warm water discharge from eight openings of 1.58 m diameter, at the location of approximately 300 m eastward of the intake. Its hydraulic studies and model experiments and its construction are reported. (Wakatsuki, Y.)

  9. Hydrogen behaviour and mitigation in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Della Loggia, E.

    1992-01-01

    The Commission of the European Communities (CEC) and the International Atomic Energy Agency (IAEA), within the framework of their safety research activities, initiated and arranged a series of specialist meetings and research contracts on hydrogen behaviour and control. The result of this work is summarized in a report jointly prepared by the two international organizations entitled 'Hydrogen in water-cooled nuclear power reactors'. Independently, the Kurchatov Atomic Energy Institute organized a workshop on the hydrogen issue in Sukhumi, USSR, with CEC and IAEA cooperation. Commonly expressed views have emerged and recommendations were formulated to organize the subsequent seminar/workshop concentrating mainly on the most recent research and analytical projects and findings related to the hydrogen behaviour, and-most importantly-on the practical approaches and engineering solutions to the hydrogen control and mitigation. The seminar/workshop, therefore, addressed the 'theory and practice' aspects of the hydrogen issue. The workshop was structured in the following sessions: combustible gas production; hydrogen distribution; combustion phenomena; combustion effects and threats; and detection and migration

  10. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    International Nuclear Information System (INIS)

    Hutter, E.; Pardini, J.A.

    1977-01-01

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads. 3 claims, 6 figures

  11. Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

    International Nuclear Information System (INIS)

    Pradal, L.; Berte, M.; Chiarelli, C.

    1985-01-01

    The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

  12. Rehabilitation of two natural draught cooling towers at Grohnde 1300 MW nuclear power station, taking into account a completely new concept

    International Nuclear Information System (INIS)

    Schwickert, M.; Meyer, V.

    1992-01-01

    The natural draught cooling towers for Grohnde Nuclear Power Station were completed in 1983. During the operating period from 1984 to 1990, partial areas of these cooling tower structures collapsed. A combination of high performance cooling installations with so-called spray screens were offered for the necessary rehabilitation. Since rehabilitation of both cooling towers had to be carried out during the operation of the power station, parts of the surfaces of the cooling towers were closed off in order to be able to carry out the difficult installation of the structures. Acceptance measurements have confirmed the thermodynamic calculations. (orig.) [de

  13. Assessment of missiles generated by pressure component failure and its application to recent gas-cooled nuclear plant design

    International Nuclear Information System (INIS)

    Tulacz, J.; Smith, R.E.

    1980-01-01

    Methods for establishing characteristics of missiles following pressure barrier rupture have been reviewed in order to enable evaluation of structural response to missile impact and to aid the design of barriers to protect essential plant on gas cooled nuclear plant against unacceptable damage from missile impact. Methods for determining structural response of concrete barriers to missile impact have been reviewed and some methods used for assessing the adequacy of steel barriers on gas-cooled nuclear plant have been described. The possibility of making an incredibility case for some of the worst missiles based on probability arguments is briefly discussed. It is shown that there may be scope for such arguments but there are difficulties in quantifying some of the probability factors. (U.K.)

  14. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhi Shang, E-mail: zhi.shang@stfc.ac.uk [Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.com [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2011-11-15

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In a vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in a horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-{epsilon} turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculation region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  15. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    International Nuclear Information System (INIS)

    Shang, Zhi; Lo, Simon

    2009-01-01

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-ε turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculating region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. (author)

  16. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  17. Macrophytes in the cooling ponds of Ukrainian nuclear and thermal power plants

    International Nuclear Information System (INIS)

    D'yachenko, T.N.

    2013-01-01

    Attention is focused at the macrophytes role in the functioning of the natural-technological cooling ponds ecosystems, at the features of aquatic plants and station water supply system interaction. It was considered the degree of macrophytes scrutiny and it was pointed out the necessity of monitoring and controlling their condition in the cooling ponds of Ukrainian power plants.

  18. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  19. Gamma spectrometric characterization of short cooling time nuclear spent fuels using hemispheric CdZnTe detectors

    CERN Document Server

    Lebrun, A; Szabó, J L; Arenas-Carrasco, J; Arlt, R; Dubreuil, A; Esmailpur-Kazerouni, K

    2000-01-01

    After years of cooling, nuclear spent fuel gamma emissions are mainly due to caesium isotopes which are emitters at 605, 662 and 796-801 keV. Extensive work has been done on such fuels using various CdTe or CdZnTe probes. When fuels have to be measured after short cooling time (during NPP outage) the spectrum is much more complex due to the important contributions of niobium and zirconium in the 700 keV range. For the first time in a nuclear power plant, four spent fuels of the Kozloduy VVER reactor no 4 were measured during outage, 37 days after shutdown of the reactor. In such conditions, good resolution is of particular interest, so a 20 mm sup 3 hemispheric crystal was used with a resolution better than 7 keV at 662 keV. This paper presents the experimental device and analyzes the results which show that CdZnTe commercially available detectors enabled us to perform a semi-quantitative determination of the burn-up after a short cooling time. In addition, it is discussed how a burn-up evolution code (CESAR)...

  20. Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leonardi, Tauna [Schlumberger, 14910 Airline Rd., Rosharon, TX 77583 (United States)]. E-mail: Tleonardi@slb.com; Ishii, Mamoru [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)]. E-mail: Ishii@ecn.purdue.edu

    2006-09-15

    Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.

  1. Consideration of inflation in comparing cooling systems for a nuclear power station

    International Nuclear Information System (INIS)

    Sullivan, W.G.; Fell, H.A.

    1976-01-01

    Based only on the initial capital outlay for each cooling system, natural-draft cooling towers were $7.828 x 10 6 less expensive than once-through cooling. This margin favoring natural-draft cooling dropped to only $2.24 x 10 6 when the present-worth of after-tax cash flow was evaluated in terms of constant-worth dollars. Annual estimates of after-tax cash flow took into account various costs that were either responsive or non-responsive to inflation, the method of loan repayment, and local and federal taxes. It is believed that the tabular procedure presented in the article has widespread applicability to other economic analyses involving the presence of inflation. Its simplicity and ease of understanding should be appealing to decision makers

  2. Nuclear reactor core support incorporating also a cooling fluid flow system

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    A description is given of a core bearing plate with several modular intake units having cooling fluid intake openings on their lower extensions, and on their upper ends located above the bearing plate, at least one fuel assembly which is thus in communication with the area under the bearing plate through the modular intake unit. The means for introducing the cooling fluid into the reactor vessel area are located under the bearing plate. The lower ends of the modular intake have ribs arranged essentially on a plane and join together with openings provided between the seals, in such a manner that the ribs form a barrier. The cooling fluid intake openings are located above this barrier, so that the cooling fluid is compelled to cross it before penetrating into the modular intake units [fr

  3. Process fluid cooling system

    International Nuclear Information System (INIS)

    Farquhar, N.G.; Schwab, J.A.

    1977-01-01

    A system of heat exchangers is disclosed for cooling process fluids. The system is particularly applicable to cooling steam generator blowdown fluid in a nuclear plant prior to chemical purification of the fluid in which it minimizes the potential of boiling of the plant cooling water which cools the blowdown fluid

  4. Eye-flukes in fish, living in cooling water from nuclear power plants

    International Nuclear Information System (INIS)

    Hoeglund, J.; Thulin, J.

    1988-01-01

    We report here on the effects of raised water temperature on the prevalence, mean infrapopulation density and consequences of eye-flukes in fish. The study was mainly performed in the Biotest basin situated 120 km north of Stockholm, Sweden. This 1 km 2 basin is an enclosed brackish water (5 permillage) area receiving heated (about 8 degree C) cooling water (90 m 3 /s) from Forsmark nuclear power station. Both morphological and experimental studies of the parasite larvae of sampled fish indicate that we are dealing with four strains of Diplostomum, two of which occur in perch and the other two in roach. Since taxonomic revisions are under hand elsewhere we prefer to name these strains Diplostomum sp1-4. Metacercariae of D.sp 1 were found between the retina and sclera in the eye of perch while that of D.sp 4 were found in the eye-lens of roach in over 90 % of fish examined. Metacercariae of the other two D.-species and of Tylodelphys clavata and Cotylurus sp were found at lower frequencies. Cercariae of Diplostomum spp were found to develop from sporocysts in snails of the genus Lymnaea. The period of cercarial shedding starts about one month earlier and is also prolonged in the Biotest basin compared to the reference locality. The infection procedure, however, is the same in both areas. During experimental infections with cercariae on yearlings of bleak we found a distinct correlation between an increased fry mortality and an increased cercariae density, a connection which was strengthened at increased water temperature. Furthermore, the results indicate that the defence mechanisms of the fish respond slower towards infections with Diplostomum spp than that is known to be the case with bacterial infections. The speed with which the metacercariae accumulate in the eye of the fish is higher in the Biotest basin than in the reference locality. In spite of this, the mean infrapopulation density of metacercariae in older fish is not higher here than in the reference

  5. A licence to discharge cooling waters in tidal rivers, examplified by the 'Nuclear Power Station Unterweser'

    International Nuclear Information System (INIS)

    Kunz, H.

    1976-01-01

    Illustrated by the example of the lower Weser, aspects for automatic control, supervision measurements, and measurements for the securing of evidence, all in connection with cooling water discharges, are presented. The particularities of tidal rivers and the conditions for measuring systems resulting therefrom are explained. The cooling water discharge of the Kernkraftwerk Unterweser has been assigned an extensive measurement system for the automatic compilation of hydrologic data. The measurement systems design, the measurement stations, and the central station are described. (orig.) [de

  6. RETRAN-02 analysis of upper head cooling during controlled natural circulation cooldown of Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Fujita, N.; Helrich, R.E.; Bergeron, P.A.

    1982-01-01

    RETRAN-02 is particularly well-suited for investigating the fluid conditions in the upper head during a natural circulation cooldown. The RETRAN input model was developed with four basic objectives: (1) accurate description of the upper head cooling mechanisms; (2) proper simulation of natural circulation; (3) respresentations of operator actions required to proceed from full-power to shutdown-cooling-system conditions using both automatic and manual controls; and (4) reduction of the computer cost of simulating this evolution of approximately 10-hour duration. The response of the upper head fluid temperature calculated by RETRAN was in close agreement with measured data obtained from a natural circulation cooldown experiment performed for the Connecticut Yankee Plant, whose design is very similar to the Yankee Nuclear Power Station

  7. Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

    International Nuclear Information System (INIS)

    Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

    2012-01-01

    Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

  8. Current status of nuclear power generation in Japan and directions in water cooled reactor technology development

    International Nuclear Information System (INIS)

    Miwa, T.

    1991-01-01

    Electric power demand aspects and current status of nuclear power generation in Japan are outlined. Although the future plan for nuclear power generation has not been determined yet the Japanese nuclear research centers and institutes are investigating and developing some projects on the next generation of light water reactors and other types of reactors. The paper describes these main activities

  9. Dresden Nuclear Power Station, Unit No. 1: Primary cooling system chemical decontamination: Draft environmental statement (Docket No. 50-10)

    International Nuclear Information System (INIS)

    1980-05-01

    The staff has considered the environmental impact and economic costs of the proposed primary cooling system chemical decontamination at Dresden Nuclear Power Station, Unit 1. The staff has focused this statement on the occupational radiation exposure associated with the proposed Unit 1 decontamination program, on alternatives to chemical decontamination, and on the environmental impact of the disposal of the solid radioactive waste generated by this decontamination. The staff has concluded that the proposed decontamination will not significantly affect the quality of the human environment. Furthermore, any impacts from the decontamination program are outweighed by its benefits. 2 figs., 7 tabs

  10. Application of the failure modes and effects analysis technique to the emergency cooling system of an experimental nuclear power plant

    International Nuclear Information System (INIS)

    Conceicao Junior, Osmar; Silva, Antonio Teixeira e

    2009-01-01

    This study consists on the application of the failure modes and effects analysis (FMEA), a hazard identification and a risk assessment technique, to the emergency cooling system (ECS), of an experimental nuclear power plant. The choice of this technique was due to its detailed analysis of each component of the system, enabling the identification of all possible ways of failure and its related consequences (in order of importance), allowing the designer to improve the system, maximizing its security and reliability. Through the application of this methodology, it could be observed that the ECS is an intrinsically safe system, in spite of the modifications proposed. (author)

  11. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  12. Evaluation of Control and Protection System for Loss of Electrical Power Supply System of Water-Cooling Nuclear Power Plant

    International Nuclear Information System (INIS)

    Suhaemi, Tjipta; Djen Djen; Setyono; Jambiar, Riswan; Rozali, Bang; Setyo P, Dwi; Tjahyono, Hendro

    2000-01-01

    Evaluation of control and protection system for loss of electrical power supply system of water-cooled nuclear power plant has been done. The loss of electrical power supply. The accident covered the loss of external electrical load and loss of ac power to the station auxiliaries. It is analysed by studying and observing the mechanism of electrical power system and mechanism of related control and protection system. The are two condition used in the evaluation i e without turbine trip and with turbine trip. From the evaluation it is concluded that the control and protection system can handled the failure caused by the loss of electrical power system

  13. Application of the failure modes and effects analysis technique to theemergency cooling system of an experimental nuclear power plant

    International Nuclear Information System (INIS)

    Conceicao Junior, Osmar

    2009-01-01

    This study consists on the application of the Failure Modes and EffectsAnalysis (FMEA), a hazard identification and a risk assessment technique, tothe Emergency Cooling System (ECS) of an experimental nuclear power plant,which is responsible for mitigating the consequences of an eventual loss ofcoolant accident on the Pressurized Water Reactor (PWR). Such analysisintends to identify possible weaknesses on the design of the system andpropose some improvements in order to maximize its reliability. To achievethis goal a detailed study of the system was carried on (through itstechnical documentation), the correspondent reliability block diagram wasobtained, the FMEA analysis was executed and, finally, some suggestions werepresented. (author)

  14. Development and computational simulation of thermoelectric electromagnetic pumps for controlling the fluid flow in liquid metal cooled space nuclear reactors

    International Nuclear Information System (INIS)

    Borges, E.M.

    1991-01-01

    Thermoelectric Electromagnetic (TEEM) Pumps can be used for controlling the fluid flow in the primary and secondary circuits of liquid metal cooled space nuclear reactor. In order to simulate and to evaluate the pumps performance, in steady-state, the computer program BEMTE has been developed to study the main operational parameters and to determine the system actuation point, for a given reactor operating power. The results for each stage of the program were satisfactory, compared to experimental data. The program shows to be adequate for the design and simulating of direct current electromagnetic pumps. (author)

  15. 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)

  16. System for cooling the upper wall of a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Pailla, Henri; Schaller, Karl; Vidard, Michel.

    1974-01-01

    A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

  17. Process and device for cooling of nuclear reactor fuel elements enclosed in a transport container

    International Nuclear Information System (INIS)

    Stiefel, M.

    1986-01-01

    In order to remove the post-decay heat of the fuel elements contained in them, transport containers for burnt-up fuel elements can be connected to a water cooling circuit. In order to avoid thermal shocks, a tenside forming foam and air are introduced into the cooling circuit before its entry into the transport container in the direction of flow. The tenside and air continue to be supplied until the temperature inside the transport container has fallen below the temperature at which the foam is destroyed. By adding tenside and air, a two phase mixture is produced, which foams greatly when it enters the transport container and which cools the fuel elements so as to protect them.(orig./HP) [de

  18. Method of avoiding hazards resulting from accidents in water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Dorner, S.; Schretzmann, K.; Schumacher, G.

    1984-01-01

    In water-cooled reactors, e.g. BWRs and PWRs, elemental hydrogen is released by hydrolysis (in-leakage). In case of an accident in these reactors or at emergency cooling of e.g., a gas-cooled reactor with water additional hydrogen is produced by chemical reactions of the water with the cladding material. In order to prevent hydrogen pressurizing and the formation of a detonating gas mixture, dry powder containers are provided for in the endangered compartments of the reactor. In case of danger powdered CuO, MnO 2 , Fe 2 O 3 , or CdO, the oxygen content of which recombines with the hydrogen, is ejected from them. In addition, an extinguishing substance with an anticatalytic resp. inhibition effect and/or an inert gas of the group N 2 , He, Ar, CO 2 may be admixed to the powder resp. powder mixture. (orig./PW)

  19. Future nuclear systems, Astrid, an option for the fourth generation: preparing the future of nuclear energy, sustainably optimising resources, defining technological options, sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ter Minassian, Vahe

    2016-01-01

    Energy independence and security of supplies, improved safety standards, sustainably optimised material management, minimal waste production - all without greenhouse gas emissions. These are the Generation IV International Forum specifications for nuclear energy of the future. The CEA is responsible for designing Astrid, an integrated technology demonstrator for the 4. generation of sodium-cooled fast reactors, in accordance with the French Sustainable Nuclear Materials and Waste Management Act of June 28, 2006, and funded as part of the Investments for the Future programme enacted by the French parliament in 2010. Energy management - a vital need and a factor of economic growth - is a major challenge for the world of tomorrow. The nuclear industry has significant advantages in this regard, although it faces safety, resource sustainability, and waste management issues that must be met through continuing technological innovation. Fast reactors are also of interest to the nuclear industry because their recycling capability would solve a number of problems related to the stockpiles of uranium and plutonium. After the resumption of R and D work with EDF and AREVA in 2006, the Astrid design studies began in 2010. The CEA, as owner and contracting authority for this programme, is now in a position to define the broad outlines of the demonstrator 4. generation reactor that could be commissioned during the next decade. A sodium-cooled fast reactor (SFR) operates in the same way as a conventional nuclear reactor: fission reactions in the atoms of fuel in the core generate heat, which is conveyed to a turbine generator to produce electricity. In the context of 4. generation technology, SFRs represent an innovative solution for optimising the use of raw materials as well as for enhancing safety. Here are a few ideas advanced by the CEA. (authors)

  20. Modeling the high-temperature gas-cooled reactor process heat plant: a nuclear to chemical conversion process

    International Nuclear Information System (INIS)

    Pfremmer, R.D.; Openshaw, F.L.

    1982-05-01

    The high-temperature heat available from the High-Temperature Gas-Cooled Reactor (HTGR) makes it suitable for many process applications. One of these applications is a large-scale energy production plant where nuclear energy is converted into chemical energy and stored for industrial or utility applications. This concept combines presently available nuclear HTGR technology and energy conversion chemical technology. The design of this complex plant involves questions of interacting plant dynamics and overall plant control. This paper discusses how these questions were answered with the aid of a hybrid computer model that was developed within the time-frame of the conceptual design studies. A brief discussion is given of the generally good operability shown for the plant and of the specific potential problems and their anticipated solution. The paper stresses the advantages of providing this information in the earliest conceptual phases of the design

  1. Study of the oxidation mechanisms between impurities and surfaces applied to the future gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Duval, A.

    2010-01-01

    Inconel 617, main candidate for the heat exchangers of the gas-cooled next generation of nuclear reactors has been investigated. Two different problems occurring in the cooling system splits the study into two parts. Oxidizing impurities contained in the coolant can cause severe corrosion at 850 C. Radioactive impurities, coming from the fission reaction of the core can, in another hand contaminate the cooling loop and cause radioprotection problem for the maintenance and dismantling operations. Firstly, oxidizing gas partial pressure influence on oxidation of IN 617 at 850 C was investigated varying oxygen and water vapour partial pressure between 1.10 -5 mbar and 200 mbar. Oxide layers were characterized using XPS, SEM, EDX, GD-OES, XRD. Influence of partial pressure on layers structure and composition was determined. Effect of water vapour and partial pressure on growth mechanisms were also investigated. The second part of this study is focused on diffusion of Ag, stable isotope of Ag-110m in IN617 alloy and in the oxide layer forming at its surface at 850 C. Concentration profiles were obtained by GD-OES calibrated analysis. Diffusion coefficient could be obtained from these diffusion profiles: volume diffusion and grain boundary diffusion coefficients for the diffusion in the alloy, and an apparent diffusion coefficient for the diffusion in the oxide, due to the porosity of the structure. (author) [fr

  2. Control room conceptual design of nuclear power plant with multiple modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Jia Qianqian; Qu Ronghong; Zhang Liangju

    2014-01-01

    A conceptual design of the control room layout for the nuclear power plant with multiple modular high temperature gas-cooled reactors has been developed. The modular high temperature gas-cooled reactors may need to be grouped to produce as much energy as a utility demands to realize the economic efficiency. There are many differences between the multi-modular plant and the current NPPs in the control room. These differences may include the staffing level, the human-machine interface design, the operation mode, etc. The potential challenges of the human factor engineering (HFE) in the control room of the multi-modular plant are analyzed, including the operation workload of the multi-modular tasks, how to help the crew to keep situation awareness of all modules, and how to support team work, the control of shared system between modules, etc. A concept design of control room for the multi-modular plant is presented based on the design aspect of HTR-PM (High temperature gas-cooled reactor pebble bed module). HFE issues are considered in the conceptual design of control room for the multi-modular plant and some design strategies are presented. As a novel conceptual design, verifications and validations are needed, and focus of further work is sketch out. (author)

  3. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  4. Phytoplankton in the cooling pond of a nuclear fuel plant. II. Spectral analysis

    International Nuclear Information System (INIS)

    Tokarskaya, Z.B.; Smagin, A.I.; Ryzhkov, E.G.; Nikitina, L.V.

    1995-01-01

    This study continues investigations into the development dynamics of phytoplankton and hydrochemical and meteorological factors over a periods of 26 years in the cooling pond of the Mayak Production Association in the Kyzyl-Trash Lake. The aim is to evaluate the long-term oscillations in phytoplankton owing to changes in hydrochemical and meteorological factors. 6 refs., 2 figs., 1 tab

  5. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-08-01

    The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.

  6. Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

    1982-08-01

    The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste

  7. observer-based diagnostics and monitoring of vibrations in nuclear reactor core cooling system

    International Nuclear Information System (INIS)

    Siry, S.A K.

    2007-01-01

    analysis and diagnostics of vibration in industrial systems play a significant rule to prevent severe severe damages . drive shaft vibration is a complicated phenomenon composed of two independent forms of vibrations, translational and torsional. translational vibration measurements in case of the reactor core cooling system are introduced. the system under study consists of the three phase induction motor, flywheel, centrifugal pump, and two coupling between motor-flywheel, and flywheel-pump. this system structure is considered to be one where the blades are pegged into the discs fitting into the shafts. a non-linear model to simulate vibration in the reactor core cooling system will be introduced. simulation results of an operating reactor core cooling system using the actual parameters will be presented to validate the accuracy and reliability of the proposed analytical method the accuracy in analyzing the results depends on the system model. the shortcomings of the conventional model will be avoided through the use of that accurate nonlinear model which improve the simulation of the reactor core cooling system

  8. Design and neutronic investigation of the Nano fluids application to VVER-1000 nuclear reactor with dual cooled annular fuel

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Ebrahimian, M.

    2016-01-01

    Highlights: • The change in neutronic parameters to the use of nanofluid as coolant is presented. • Nanoparticle deposition on fuel clad is investigated. • Radial and axial local power peaking factors are presented. • ZrO 2 and Al 2 O 3 have the lowest rate of K eff drop off. - Abstract: Nowadays, many efforts have been made to improve the efficiency of nuclear power plants. One of which is use of the dual cooled annular fuel which is an internally and externally cooled annular fuel with many advantages in heat transfer characteristics. Another is the use of nanoparticle/water (nanofluid) as coolant. In this paper, by combining these two methods, the change in neutronic parameters of the VVER-1000 nuclear reactor core with dual cooled annular fuel attributable to the use of nanoparticle/water (nanofluid) as coolant is presented. Optimization of type and volume fraction of nanoparticles in water that affect the safety enhancement of core primary parameters is intended in this study. Reactivity change, radial and axial local power peaking factors (LPPF), and the consequence of nanoparticle deposition on fuel clad are investigated. As a result of changing the effective multiplication factor and PPF calculations for six types of nanoparticles which have been studied extensively for their heat transfer properties including Alumina, Aluminum, Copper oxide, Copper, Titania, and Zirconia with different volume fractions, it can be concluded that at low concentration (0.03 volume fraction), Zirconia and Alumina are the optimum nanoparticles for normal operation. The maximum radial and axial PPF are found to be invariant to the type of nanofluid at low volume fractions. With an increase in nanoparticle deposition thickness on the outer and inner clad, a flux and K eff depression occurred and ZrO 2 and Al 2 O 3 have the lowest rate of drop off.

  9. Benchmarking of thermalhydraulic loop models for lead-alloy-cooled advanced nuclear energy systems. Phase I: Isothermal forced convection case

    International Nuclear Information System (INIS)

    2012-06-01

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of the Fuel Cycle (WPFC) has been established to co-ordinate scientific activities regarding various existing and advanced nuclear fuel cycles, including advanced reactor systems, associated chemistry and flowsheets, development and performance of fuel and materials and accelerators and spallation targets. The WPFC has different expert groups to cover a wide range of scientific issues in the field of nuclear fuel cycle. The Task Force on Lead-Alloy-Cooled Advanced Nuclear Energy Systems (LACANES) was created in 2006 to study thermal-hydraulic characteristics of heavy liquid metal coolant loop. The objectives of the task force are to (1) validate thermal-hydraulic loop models for application to LACANES design analysis in participating organisations, by benchmarking with a set of well-characterised lead-alloy coolant loop test data, (2) establish guidelines for quantifying thermal-hydraulic modelling parameters related to friction and heat transfer by lead-alloy coolant and (3) identify specific issues, either in modelling and/or in loop testing, which need to be addressed via possible future work. Nine participants from seven different institutes participated in the first phase of the benchmark. This report provides details of the benchmark specifications, method and code characteristics and results of the preliminary study: pressure loss coefficient and Phase-I. A comparison and analysis of the results will be performed together with Phase-II

  10. Nuclear power for coexistence with nature, high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

  11. Sodium cooled research thermal reactor - a proposal to the Brazilian nuclear community

    International Nuclear Information System (INIS)

    Ishiguro, Yuji

    1996-01-01

    The nuclear community can contribute to the society in two ways: assuring reactor technologies for electric power supply and contributing to developments in other areas by application of radiations. Industrialized countries maintain intensive activities in the two senses, while in Brazil nuclear policy is not clear and opportunities of research with radiations are quite limited. It is proposed, as a way out of this situations, that the nuclear community concentrate its activities in the sense of proposing the construction of a low-power research reactor that can satisfy a majority of demands (radioisotopes, research, fast reactor) and avoid the problems of experimental fast reactors (high cost, use of Pu and HEU). (author)

  12. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Wade, D.C.; Moisseytsev, A.

    2008-01-01

    Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus

  13. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

  14. Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

    International Nuclear Information System (INIS)

    Smith, J.; Hart, R.; Lazic, L.

    2009-01-01

    Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

  15. Cationic surfactants for control of fresh- and saltwater mollusks in nuclear cooling systems

    International Nuclear Information System (INIS)

    Post, R.M.; Mallen, E.; Lehmann, F.

    1991-01-01

    One result of the release of the US Nuclear Regulatory Commission's Generic Letter 89-13, Service Water Problems Affecting Safety-Related Equipment, was the heightened awareness of the nuclear industry to the problems of macrofouling in heat exchange systems. The principal mollusk species that contribute to freshwater macrofouling problems are Asiatic Clam (southern United States) and Zebra Mussel (Great Lakes). The predominant saltwater fouling mollusks are the Blue Mussel (Pacific, northern Atlantic), Ribbed Mussel (southern Atlantic, Gulf Coast), and American Oyster (Atlantic, Gulf Coast). The nuclear community's awareness of macrofouling problems and the ineffectiveness of intermittent chlorination programs have led to the development of several chemical control technologies for eliminating macrofouling organism infestation. One technology that has proven effective for the control of macrofouling organisms is the periodic addition of a combination of two cationic charged surfactants, specifically, alkyldimethylbenzylammonium chloride (QUAT) and dodecyl guanidine hydrochloride (DGH). Experience with the cationic surfactants at several nuclear power plants is reported

  16. Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

    International Nuclear Information System (INIS)

    Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

    1987-01-01

    In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

  17. Advances in conceptual design of a gas-cooled accelerator driven system (ADS) transmutation devices to sustainable nuclear energy development

    International Nuclear Information System (INIS)

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia; Castells, Escriva; Abanades

    2011-01-01

    The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In this paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made. On the other hand, the heterogeneity of the TRISO particles inside the pebbles can not be negligible. In this paper, a study of the power density distribution inside the pebbles by means of a detailed simulation of the TRISO fuel particles and using an homogeneous composition of the fuel is addressed. (author)

  18. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  19. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  20. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  1. Passive cooling applications for nuclear power plants using pulsating steam-water heat pipes

    International Nuclear Information System (INIS)

    Aparna, J.; Chandraker, D.K.

    2015-01-01

    Gen IV reactors incorporate passive principles in their system design as an important safety philosophy. Passive safety systems use inherent physical phenomena for delivering the desired safe action without any external inputs or intrusion. The accidents in Fukushima have renewed the focus on passive self-manageable systems capable of unattended operation, for long hours even in extended station blackout (SBO) and severe accident conditions. Generally, advanced reactors use water or atmospheric air as their ultimate heat sink and employ passive principles in design for enhanced safety. This paper would be discussing the experimental results on pulsating steam water heat-pipe devices and their applications in passive cooling. (author)

  2. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  3. Water cooling in nuclear reactors by using panels of integrated circuits

    International Nuclear Information System (INIS)

    Dominique, P.; Letissier, R.

    1977-01-01

    In view of the drawbacks of wet cooling towers, EDF is searching for another approach to the problem. A self-cleaning device is now envisaged, that consists in some exchanger plates, 30 to 40m height (max. 60m) capable of being hiden in the lanscape behind high trees. The plates would be rather smooth and the air circulated by natural convection. The prototype is composed of 960 aluminium panels of integrated circuits mounted on three modules made of tubular elements working as supporting and collecting means together [fr

  4. Dynamic Thermal Loads and Cooling Requirements Calculations for V ACs System in Nuclear Fuel Processing Facilities Using Computer Aided Energy Conservation Models

    International Nuclear Information System (INIS)

    EL Fawal, M.M.; Gadalla, A.A.; Taher, B.M.

    2010-01-01

    In terms of nuclear safety, the most important function of ventilation air conditioning (VAC) systems is to maintain safe ambient conditions for components and structures important to safety inside the nuclear facility and to maintain appropriate working conditions for the plant's operating and maintenance staff. As a part of a study aimed to evaluate the performance of VAC system of the nuclear fuel cycle facility (NFCF) a computer model was developed and verified to evaluate the thermal loads and cooling requirements for different zones of fuel processing facility. The program is based on transfer function method (TFM) and it is used to calculate the dynamic heat gain by various multilayer walls constructions and windows hour by hour at any orientation of the building. The developed model was verified by comparing the obtained calculated results of the solar heat gain by a given building with the corresponding calculated values using finite difference method (FDM) and total equivalent temperature different method (TETD). As an example the developed program is used to calculate the cooling loads of the different zones of a typical nuclear fuel facility the results showed that the cooling capacities of the different cooling units of each zone of the facility meet the design requirements according to safety regulations in nuclear facilities.

  5. Cleaning device for recycling pump motor cooling system in nuclear reactor

    International Nuclear Information System (INIS)

    Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

    1996-01-01

    The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

  6. Device for monitoring radioactivity of cooling water in a nuclear reactor

    International Nuclear Information System (INIS)

    Osawa, Yasuo.

    1975-01-01

    Object: To provide means for monitoring the peak channel of γ-ray spectrum in cooling water and the time-wise attenuation value of the counts of the peak channels and capable of early detecting abnormal phenomenon with a constant reference. Structure: It is provided with a γ-ray detector, a multi-channel γ-ray spectrometer, peak determining means for determining the peak position of the spectrum from the count value of each channel of the γ-ray spectrum, a peak channel memory for memorizing the channel number of the peak channels, attenuation measurement means for measuring the attenuation value by repeatedly measuring the count value of the peak channel, an attenuation memory for memorizing the attenuation value and a variation detector for detecting the variation in radioactivity of the reactor cooling water from the count value of the peak channel and peak channel attenuation value. When a difference is detected by the variation detector, the measurement value is provided as defective value. (Kamimura, M.)

  7. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    Energy Technology Data Exchange (ETDEWEB)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

    2016-12-15

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

  8. Evaluation method of corrosive conditions in cooling systems of nuclear power plants by combined analyses of flow dynamics and corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Shunsuke [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Atomic Energy Society of Japan (AESJ) (Japan). Research Committee on Water Chemistry Standard; Naitoh, Masanori [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Atomic Energy Society of Japan (AESJ) (Japan). Computational Science and Engineering Div.; Uehara, Yasushi; Okada, Hidetoshi [Nuclear Power Engineering Corporation (NUPEC), Tokyo (Japan); Hotta, Koji [ITOCHU Techno-Solutions Corporation (Japan); Ichikawa, Ryoko [Mizuho Information and Research Inst., Inc. (Japan); Koshizuka, Seiichi [Tokyo Univ. (Japan)

    2007-03-15

    Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration, corrosion and their overlapping effects. In order to establish safe and reliable plant operation, it is necessary to predict future problems for structural materials based on combined analyses of flow dynamics and corrosion and to mitigate them before they become serious issues for plant operation. The analysis models are divided into two types. 1. Prediction models for future problems with structural materials: Distributions of oxidant concentrations along flow paths are obtained by solving water radiolysis reactions in the boiling water reactor (BWR) primary cooling water and hydrazine-oxygen reactions in the pressurized water reactor (PWR) secondary cooling water. Then, the electrochemical corrosion potential (ECP) at the point of interest is also obtained by the mixed potential model using oxidant concentration. Higher ECP enhances the possibility of intergranular stress corrosion cracking (IGSCC) in the BWR primary system, while lower ECP enhances flow accelerated corrosion (FAC) in the PWR secondary system. 2. Evaluation models of wall thinning caused by flow accelerated corrosion: The degree of wall thinning is evaluated at a location with a higher possibility of FAC occurrence, and lifetime is estimated for preventive maintenance. General features of models are reviewed in this paper and the prediction models for oxidant concentrations are briefly introduced. (orig.)

  9. Evaluation method of corrosive conditions in cooling systems of nuclear power plants by combined analyses of flow dynamics and corrosion

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Hotta, Koji; Ichikawa, Ryoko; Koshizuka, Seiichi

    2007-01-01

    Problems in major components and structural materials in nuclear power plants have often been caused by flow induced vibration, corrosion and their overlapping effects. In order to establish safe and reliable plant operation, it is necessary to predict future problems for structural materials based on combined analyses of flow dynamics and corrosion and to mitigate them before they become serious issues for plant operation. The analysis models are divided into two types. 1. Prediction models for future problems with structural materials: Distributions of oxidant concentrations along flow paths are obtained by solving water radiolysis reactions in the boiling water reactor (BWR) primary cooling water and hydrazine-oxygen reactions in the pressurized water reactor (PWR) secondary cooling water. Then, the electrochemical corrosion potential (ECP) at the point of interest is also obtained by the mixed potential model using oxidant concentration. Higher ECP enhances the possibility of intergranular stress corrosion cracking (IGSCC) in the BWR primary system, while lower ECP enhances flow accelerated corrosion (FAC) in the PWR secondary system. 2. Evaluation models of wall thinning caused by flow accelerated corrosion: The degree of wall thinning is evaluated at a location with a higher possibility of FAC occurrence, and lifetime is estimated for preventive maintenance. General features of models are reviewed in this paper and the prediction models for oxidant concentrations are briefly introduced. (orig.)

  10. The evaluation of minimum cooling period for loading of PWR spent nuclear fuel of a dual purpose metal cask

    International Nuclear Information System (INIS)

    Dho, Ho Seog; Kim, Tae Man; Cho, Chun Hyung

    2016-01-01

    Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R and D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0-4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask

  11. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  12. Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

    International Nuclear Information System (INIS)

    Feaugas, J.; Jeanjacques, M.; Peulve, J.

    1994-01-01

    G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs

  13. Calculation of mass flow and steam quality distribution on fuel elements of light-water cooled boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Hermanns, H.J.

    1977-04-01

    By the example of light-water cooled nuclear reactors, the state of the calculation methods at disposal for calculating mass flow and steam quality distribution (sub-channel analysis) is indicated. Particular regard was paid to the transport phenomena occurring in reactor fuel elements in the range of two phase flow. Experimentally determined values were compared with recalculations of these experiments with the sub-channel code COBRA; from the results of these comparing calculations, conclusions could be drawn on the suitability of this code for defined applications. Limits of reliability could be determined to some extent. Based on the experience gained and the study of individual physical model concepts, recognized as being important, a sub-channel model was drawn up and the corresponding numerical computer code (SIEWAS) worked out. Experiments made at GE could be reproduced with the code SIEWAS with sufficient accuracy. (orig.) [de

  14. Safety assessment for electricity generation failure accident of gas cooled nuclear power plant using system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2013-04-15

    The power production failure happens in the loss of coolant of the nuclear power plants (NPPs). The air ingress is a serious accident in gas cooled NPPs. The quantification of the study performed by the system dynamics (SD) method which is processed by the feedback algorithms. The Vensim software package is used for the simulation, which is performed by the Monte-Carlo method. Two kinds of considerations as the economic and safety properties are important in NPPs. The result shows the stability of the operation when the power can be decided. The maximum value of risk is the 11.77 in 43rd and the minimum value is 0.0 in several years. So, the success of the circulation of coolant is simulated by the dynamical values. (orig.)

  15. Natural Circulation in Water Cooled Nuclear Power Plants Phenomena, models, and methodology for system reliability assessments

    Energy Technology Data Exchange (ETDEWEB)

    Jose Reyes

    2005-02-14

    In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''.

  16. Natural Circulation in Water Cooled Nuclear Power Plants Phenomena, models, and methodology for system reliability assessments

    International Nuclear Information System (INIS)

    Jose Reyes

    2005-01-01

    In recent years it has been recognized that the application of passive safety systems (i.e., those whose operation takes advantage of natural forces such as convection and gravity), can contribute to simplification and potentially to improved economics of new nuclear power plant designs. In 1991 the IAEA Conference on ''The Safety of Nuclear Power: Strategy for the Future'' noted that for new plants the use of passive safety features is a desirable method of achieving simplification and increasing the reliability of the performance of essential safety functions, and should be used wherever appropriate''

  17. Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Edwards, Michael J.

    2018-01-23

    A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

  18. Phenomenology of deflagration and detonation of hydrogen-air mixtures in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Lombardi, G.

    1984-01-01

    This paper summarizes fundamentals of the flammability of the hydrogen-air mixtures and hydrogen-air containing added steam or other inerting agent. The flammability behaviour of such gaseous mixtures is described with reference to physical and chemical conditions close enough to those expected in the containment of a nuclear reactor during a LOCA

  19. Pumps of molten metal based on magnetohydrodynamicprinciple for cooling high-temperature nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Doležel, Ivo; Donátová, M.; Karban, P.; Ulrych, B.

    2009-01-01

    Roč. 85, č. 4 (2009), s. 13-15 ISSN 0033-2097 R&D Projects: GA ČR(CZ) GA102/07/0496 Institutional research plan: CEZ:AV0Z20570509 Keywords : pumps of molten metal * magnetohydrodynamic principle * nuclear reactors Subject RIV: JA - Electronics ; Optoelectronics, Electrical Engineering Impact factor: 0.196, year: 2009

  20. Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

    Science.gov (United States)

    Yacobucci, H. G.; Heestand, R. L.; Kizer, D. E.

    1973-01-01

    The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

  1. Control device for can failures of liquid cooled nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Althaus, D.; Mohm, F.; Nyhof, M.

    1974-01-01

    Checking of the fuel or breeding elements of, e.g., sodium-cooled reactors is done by detecting fission products in the coolant, with a flushing gas like argon removing the fission products from the coolant and carrying them to a detector system. In order to increase the escape rate of the fission products from the elements, these are lifted by a hoisting unit into a pit reaching down below the coolant level and are then heated. Heating is achieved by the decay heat or by an additional heating in the receiving pit which at this point is thermically insulated from the exterior. The flushing gas is blown radially into the receiving pit from below. The rising bubbles take along the fission products to a scintillation counter mounted on the head of the receiving vessel. This vessel may have double walls with coolant flowing through the interspace. (DG) [de

  2. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  3. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  4. Emergency cooling system with hot-water jet pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Reinsch, A.O.W.

    1977-01-01

    The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

  5. Conceptual study of a complementary scram system for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vanmaercke, S.; Van den Eynde, G.; Tijskens, E.; Bartosiewicz, Y.

    2009-01-01

    GEN-IV reactors promise higher safety and reliability as one of the major improvements over previous generations of reactors. To achieve that, all GEN-IV reactor concepts require two completely independent shutdown systems that rely on different operating principles. For liquid metal cooled reactors the first system is an absorber-rod based solution. The second system that by requirement should rely on another principle, is however quite a challenge to design. The second system used in current PWR reactors is to dissolve a neutron absorber, boric acid, into the primary coolant. This method cannot be used in liquid metal cooled reactors because of the high cost of cleaning the coolant. In this paper an overview of the existing literature on scram systems is given, each with their advantages and limitations. A promising new concept is also presented. This concept leads to a totally passive self activating device using small absorbing particles that flow into a dedicated channel to shutdown the reactor. The system consists of tubes filled with particles of an absorber material. During normal operation, these particles are kept above the active core by means of a metallic seal. In case of an accident, the system is activated by the temperature increase in the coolant. This leads to melting of the metal seal. The ongoing work conducted at SCK·CEN and UCL/TERM aims at assessing the reliability of this new concept both experimentally and numerically. This study is multidisciplinary as neutronic and thermal hydraulics issues are tackled. Most challenging is however the thermal hydraulics related to understanding and predicting the liberation and flow of the absorber particles during a shutdown. Simple experiments are envisaged to compare to numerical simulations using the Discrete Element Method for simulating the particles. In a later stage this will be coupled with Smoothed Particles Hydrodynamics for simulating the melting of the seal. Some preliminary experimental and

  6. Improvement in understanding of natural circulation phenomena in water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Choi, Jong-Ho; Cleveland, John; Aksan, Nusret

    2011-01-01

    Highlights: ► Phenomena influencing natural circulation in passive systems. ► Behaviour in large pools of liquid. ► Effect of non-condensable gas on condensation heat transfer. ► Behaviour of containment emergency systems. ► Natural circulation flow and pressure drop in various geometries. - Abstract: The IAEA has organized a coordinated research project (CRP) on “Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems That Utilize Natural Circulation.” Specific objectives of CRP were to (i) establish the status of knowledge: reactor start-up and operation, passive system initiation and operation, flow stability, 3-D effects, and scaling laws, (ii) investigate phenomena influencing reliability of passive natural circulation systems, (iii) review experimental databases for the phenomena, (iv) examine the ability of computer codes to predict natural circulation and related phenomena, and (v) apply methodologies for examining the reliability of passive systems. Sixteen institutes from 13 IAEA Member States have participated in this CRP. Twenty reference advanced water cooled reactor designs including evolutionary and innovative designs were selected to examine the use of natural circulation and passive systems in their designs. Twelve phenomena influencing natural circulation were identified and characterized: (1) behaviour in large pools of liquid, (2) effect of non-condensable gases on condensation heat transfer, (3) condensation on the containment structures, (4) behaviour of containment emergency systems, (5) thermo-fluid dynamics and pressure drops in various geometrical configurations, (6) natural circulation in closed loop, (7) steam liquid interaction, (8) gravity driven cooling and accumulator behaviour, (9) liquid temperature stratification, (10) behaviour of emergency heat exchangers and isolation condensers, (11) stratification and mixing of boron, and (12) core make-up tank behaviour. This paper summarizes the

  7. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  8. Water cooling coil

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, S; Ito, Y; Kazawa, Y

    1975-02-05

    Object: To provide a water cooling coil in a toroidal nuclear fusion device, in which coil is formed into a small-size in section so as not to increase dimensions, weight or the like of machineries including the coil. Structure: A conductor arranged as an outermost layer of a multiple-wind water cooling coil comprises a hollow conductor, which is directly cooled by fluid, and as a consequence, a solid conductor disposed interiorly thereof is cooled indirectly.

  9. UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

    Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

  10. Nuclear power station with a water-cooled reactor pressure vessel

    International Nuclear Information System (INIS)

    Hoffmann, R.; Brunner, G.; Jost, N.

    1987-01-01

    Nuclear radiation produces radiolysis gases, which are undesirable for corrosion and oxyhydrogen gas reasons. To limit the proportion of this radiolysis gas, the invention provides that catalytic surfaces should be introduced into the primary circuit, to produce recombination of hydrogen and oxygen. These surfaces can be accommodated in the upper part of the reactor pressure vessel. The live steam screen can also have a catalytic surface. (orig./HP) [de

  11. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    International Nuclear Information System (INIS)

    Beck, J.M.; Collins, J.W.; Garcia, C.B.; Pincock, L.F.

    2010-01-01

    High Temperature Gas Reactors (HTGR) have been designed and operated throughout the world over the past five decades. These seven HTGRs are varied in size, outlet temperature, primary fluid, and purpose. However, there is much the Next Generation Nuclear Plant (NGNP) has learned and can learn from these experiences. This report captures these various experiences and documents the lessons learned according to the physical NGNP hardware (i.e., systems, subsystems, and components) affected thereby.

  12. Structure of liquid metal cooled nuclear reactor with loops and steady vessel

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    This structure comprises, in a vessel containing liquid metal, a nuclear core steadied on an alimentation diagrid and external loops comprising heat exchanger and reinjection pump of sodium in the diagrid. The vessel has the bottom resting on the concrete surround with a thermal stratification of the sodium between the bottom and the diagrid. This disposition has for advantage to allow a vertical connection of the sodium reinjection channel. This channel is contained in a metal sheath with a sliding leak tightness [fr

  13. Hydrology of lake Druksiai, the cooling pond of the Ignalina nuclear plant

    International Nuclear Information System (INIS)

    Lasinskas, M.

    1994-01-01

    The article presents the data on the water balance in lake Druksiai and evaluation of data on surface evaporation, surface temperature distribution, hydrothermal behaviour and water level variations under the influence of the Ignalina Nuclear Power Plant. A list is supplied of reports on the studies of hydrology in lake Druksiai, as well as of books and journal articles containing the results of studies of the lake. (author). 51 refs., 3 tabs., 10 figs

  14. Low-temperature dynamic nuclear polarization with helium-cooled samples and nitrogen-driven magic-angle spinning.

    Science.gov (United States)

    Thurber, Kent; Tycko, Robert

    2016-03-01

    We describe novel instrumentation for low-temperature solid state nuclear magnetic resonance (NMR) with dynamic nuclear polarization (DNP) and magic-angle spinning (MAS), focusing on aspects of this instrumentation that have not been described in detail in previous publications. We characterize the performance of an extended interaction oscillator (EIO) microwave source, operating near 264 GHz with 1.5 W output power, which we use in conjunction with a quasi-optical microwave polarizing system and a MAS NMR probe that employs liquid helium for sample cooling and nitrogen gas for sample spinning. Enhancement factors for cross-polarized (13)C NMR signals in the 100-200 range are demonstrated with DNP at 25K. The dependences of signal amplitudes on sample temperature, as well as microwave power, polarization, and frequency, are presented. We show that sample temperatures below 30K can be achieved with helium consumption rates below 1.3 l/h. To illustrate potential applications of this instrumentation in structural studies of biochemical systems, we compare results from low-temperature DNP experiments on a calmodulin-binding peptide in its free and bound states. Published by Elsevier Inc.

  15. ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

    International Nuclear Information System (INIS)

    Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

    1977-01-01

    ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

  16. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    International Nuclear Information System (INIS)

    Mella, R.; Wenman, M.R.

    2013-01-01

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  17. 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)

  18. Emergency cooling system for a nuclear reactor in a closed gas turbine plant

    International Nuclear Information System (INIS)

    Frutschi, H.U.

    1974-01-01

    In undisturbed operation of the closed gas turbine plant with compressor stages, reactor, and turbine, a compressor stage driven by a separate motor is following with reduced power. The power input this way is so small that the working medium is just blown through without pressure increase. The compressor stage is connected with the reactor by means of a reactor feedback pipe with an additional cooler and with the other compressor stages by means of a recuperator in the pipe between these and the turbine. In case of emergency cooling, e.g. after the rupture of a pipe with decreasing pressure of the working medium, the feedback pipe is closed short and the additional compressor stage is brought to higher power. It serves as a coolant blower and transfers the necessary amount of working medium to the reactor. The compressor stage is controlled at a constant torque, so that the heat removal from the reactor is adapted to the conditions of the accident. (DG) [de

  19. The nuclear design optimization of a Pb-Bi alloy cooled transmuter, PEACER-300

    International Nuclear Information System (INIS)

    Lim, Jae-Yong; Kim, Myung-Hyun

    2006-01-01

    A core design of lead-bismuth cooled fast reactor, PEACER-300 has been investigated to maximize its transmutation capability within safety criteria. Transmutation of minor actinide under a closed recycling was analyzed with assumption on decontamination factors in pyro-reprocessing plant data at reasonably high values. To acquire high transmutation performance, feed fuel composition, P/D ratio, active core height and fuel cycle strategy were changed. For preventing the fuel meting and guaranteeing long plant life-time, the number of fuel assembly array and normal operation temperature were decided. The optimized design parameter were chosen as of a flat core shape with 50 cm of active core height and 5 m core diameter, loaded with 17 x 17 arrayed fuel assemblies. A pitch to diameter ratio is 2.2, operating coolant temperature range is 300 deg. C to 400 deg. C, and core consists of 3 different enrichment zones with one year cycle length. Performance of designed core showed a high transmutation capability with support ratio of 2.085, large negative temperature feedback coefficients, and sufficient shutdown margin with 28 B 4 C control assemblies. (authors)

  20. Material development for gas-cooled high temperature reactors for the production of nuclear process heat

    International Nuclear Information System (INIS)

    Nickel, H.

    1977-04-01

    In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

  1. Disordered nuclear pasta, magnetic field decay, and crust cooling in neutron stars

    OpenAIRE

    Horowitz, C. J.; Berry, D. K.; Briggs, C. M.; Caplan, M. E.; Cumming, A.; Schneider, A. S.

    2014-01-01

    Nuclear pasta, with non-spherical shapes, is expected near the base of the crust in neutron stars. Large scale molecular dynamics simulations of pasta show long lived topological defects that could increase electron scattering and reduce both the thermal and electrical conductivities. We model a possible low conductivity pasta layer by increasing an impurity parameter Q_{imp}. Predictions of light curves for the low mass X-ray binary MXB 1659-29, assuming a large Q_{imp}, find continued late ...

  2. Hyperfine enhanced nuclear magnetic cooling of PrCu6 pressed in indium

    International Nuclear Information System (INIS)

    Lewis, J.D.

    1976-06-01

    Pressing of PrCu 6 in In was studied as part of the study of the preparation of nuclear coolants. Problems of low thermal conductivity through the bulk sample appear to be solved using this method. Indium was shown to be useful as a matrix ''glue'' owing to its low heat capacity and desirable mechanical properties. The eddy current heating increase observed in the PrCu 6 --In sample was not excessive. The study also showed that the quality of demagnetization samples depends on preparation techniques. 12 figures, 28 references

  3. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    Costa, Antonio Carlos Lopes da; Machado, Luiz; Koury, Ricardo Nicolau Nassar; Passos, Julio Cesar

    2009-01-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  4. Industrial space heating and cooling from stored spent nuclear power plant fuel

    International Nuclear Information System (INIS)

    Shaver, B.O.; Doman, J.W.

    1980-01-01

    Projections by the Department of Energy indicate that some 5800 metric tons of spent fuel from nuclear power reactors are now in storage and that some 33000 metric tons are expected to be in storage in 1990. The bulk of the spent fuel is currently stored in water-filled basins at the reactor sites from which the material was discharged. The thermal energy in the fuel is dissipated to atmospheres via a pumped water-to-air heat exchanger system. This paper describes a feasibility study of potential methods for the use of the heat. Also, potential applications of heat recovery systems at larger AFR storage facilities were investigated

  5. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  6. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-04-01

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  7. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-04-01

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  8. Purification of cooling water for nuclear reactors using ion exchangers; Preciscavanje vode za hladjenje nuklearnih reaktora pomocu neorganskih jonoizmenjivaca

    Energy Technology Data Exchange (ETDEWEB)

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

    1969-07-01

    Zirconiumphosphate, zirconiumoxide and natural magnetite as inorganic substances with favourable adsorption properties were the subject of investigations dealing with problems of water purification for nuclear rector cooling. Study on adsorption of impurities form reactor water to 300 deg C and 100 Atm was done by specially constructed autoclaves. On the other hand, a pre-project covering a laboratory plant for investigation of inorganic ion exchangers under real dynamic conditions is given. In order to obtain necessary data on the basis of which techno-economical analyses regarding utilization of zirconiumphosphate, zirconiumoxide and magnetite for water purification is cooling the reactors types BWR and PWR, could be performed, systematic investigations of physical and chemical properties of these substances were commenced. Equilibrium constants have been determined for adsorption processes at different pH values, as well as under various temperatures. Obtained equilibrium constants were used for calculation of thermodynamic quantities {delta}H, {delta}G and {delta}S (author) [Serbo-Croat] Cirkonijumfosfat, cirkonijumoksid i prirodni magnet, kao neorganski materijali sa pogodnim adsorpcionim osobinama, bili su predmet istrazivanja vezanih za probleme preciscavanja vode za hladjenje nuklearnih reaktora. Izucavanje adsorpcije necistoca iz reaktorske vode do 300 deg C i 100 Atm vrseno je pomocu specijalno konstruisanog autoklava, a za ispitivanje neorganskih jonoizmenjivaca pri realnim dinamickim uslovima dat je idejni projekt jednog laboratorijskog postrojenja. Za dobijanje potrebnih podataka, na osnovu kojih se mogu napraviti tehno-ekonomske analize o koriscenju cirkonijumfosfata, cirkoijumoksida i magnetita za preciscavanje vode za hladjenje reaktora tipa BWR i PWR, zapoceto je sa sistematskim proucavanjima fizickih i hemijskih osobina pomenutih materijala, odredjivane su konstante ravnoteze za procese adsorpcije pri razlicitim pH vrednostima, kao i na razlicitim

  9. SPLOSH II: A dynamics programme for nuclear - thermal - hydrodynamic behaviour of water-cooled reactors

    International Nuclear Information System (INIS)

    Moxon, D.

    1966-01-01

    A dynamics code is described that solves the two-group neutron diffusion equations simultaneously with the thermal and the hydraulic equations for an average channel of a water-cooled reactor. Other reactor channels can be represented as 'slaves', which have no feedback to the average channel. The fission power at any axial station in a slave channel is related to that in the average by prescribed time-dependent factors, and the hydraulic flow is determined from pressure-drop requirements dictated by the performance of the average channel. A finite difference model of the fuel element and can represents the behaviour of the fuel temperatures and surface heat flux. The representation of the hydraulic circuit has been made sufficiently general that the code is applicable to B.W.R., P.W.R. and pressure tube reactor designs. The code can be used to study transients resulting from imposed time variations in coolant flow, inlet enthalpy, system pressure, electrical torque supplied to the circulating pumps, (or alternatively, the angular velocity of the pump rotors,) moderator height, frictional resistances simulating blockages and control rod and fuel element insertions. The harmonic response can be obtained by injecting sinusoidal time variations until the starting transient has been damped out. Output includes axial distributions of the neutron fluxes, heat flux, coolant density and temperature, burn-but margin, and the fuel and can temperatures in both the average and the slave channels. The code was originally written in FORTRAN II for use on the IBM 7090. Computing times vary greatly with the problem and the desired accuracy but experience has shown that a computing time which is slower than real time by a factor thirty is adequate for a wide range of cases. The code has recently been converted to S2 and EGTRAN for use on the IBM 7030 and the English Electric Leo Marconi KDF 9 computers. (author)

  10. Radiation protection training for personnel at light-water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    Section 19.12 Instructions to Workers, of 10 CFR Part 19, Notices, Instructions, and Reports to Workers; Inspections, requires that individuals be given instruction in radiation protection that is commensurate with the potential radiation protection problems they may encounter in restricted areas as defined in para. 19.3(e) of 10 CFR Part 19. Para. 20.1(c) of 10 CFR Part 20, Standards for Protection Against Radiation, states that occupational radiation exposure should be kept as low as is reasonably achievable (ALARA). Appropriate training is an essential aspect of an ALARA program. This guide describes a radiation protection training program consistent with the ALARA objective and acceptable to the NRC staff for meeting the training requirements of 10 CFR Part 19 with respect to individuals that enter restricted areas at nuclear power plants

  11. The use of reliability analysis techniques applied to nuclear power station emergency core cooling systems

    International Nuclear Information System (INIS)

    Danielsen, A.; Snaith, E.R.

    1975-01-01

    A reliability investigation carried out by the Safety and Reliability Services of the UKAEA, and the SSEB, of the essential system/reactor coolant system for a large nuclear power station is described. In AGR type reactors, after all reactor shutdown conditions, it is necessary to restore forced gas circulation and sufficient boiler feed to maintain the heat removal capacity of the boilers. The coolant requirements are provided by several independent mechanical systems of primary coolant fans, feedwater pumps, and valves integrated with electrical power sources, switchgear, and automatic control equipment. Reliability is treated as one aspect of system performance and quantified in terms of failure to meet a specific objective. Based on the reliability performance of the constituent components the optimum system configuration is determined together with the preferred plant operating procedures and maintenance requirements. (author)

  12. Distribution of silica species in cooling water system in nuclear power station

    International Nuclear Information System (INIS)

    Akiba, Kenichi; Onozuka, Teruo; Shindo, Manabu.

    1995-01-01

    Distribution of silica species was examined by spectrophotometric method based on the formation of molybdosilicic acid species. Ultra-microamounts of ionic (reactive) silica were determined by collection of silicomolybdenum blue compound on a nitrocellulose membrane filter. Total concentrations of silica including nonionic (polymer and colloidal) species were also determined after decomposition of unreactive silica in alkali solutions. Water in the nuclear reactor (Onagawa BWR No.1) contained high concentration of silica (∼600 ppb) and ionic silica was found to be predominant (∼90%). In condensate system, silica contents were of a lower level (2-6 ppb), but the ionic silica contents were comparable to others (20-60%). The silica species appear to be brought and accumulated in the reactor from the condensate system, and then the silica species change to ionic species under high pressure and high temperature. (author)

  13. Distribution of silica species in cooling water system in nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Akiba, Kenichi [Tohoku Univ., Sendai (Japan). Inst. for Advanced Materials Processing; Onozuka, Teruo; Shindo, Manabu

    1995-12-01

    Distribution of silica species was examined by spectrophotometric method based on the formation of molybdosilicic acid species. Ultra-microamounts of ionic (reactive) silica were determined by collection of silicomolybdenum blue compound on a nitrocellulose membrane filter. Total concentrations of silica including nonionic (polymer and colloidal) species were also determined after decomposition of unreactive silica in alkali solutions. Water in the nuclear reactor (Onagawa BWR No.1) contained high concentration of silica ({approx}600 ppb) and ionic silica was found to be predominant ({approx}90%). In condensate system, silica contents were of a lower level (2-6 ppb), but the ionic silica contents were comparable to others (20-60%). The silica species appear to be brought and accumulated in the reactor from the condensate system, and then the silica species change to ionic species under high pressure and high temperature. (author).

  14. System for bearing a nuclear reactor vessel cooled by liquid metal

    International Nuclear Information System (INIS)

    Mahe, A.; Jullien, G.

    1976-01-01

    The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

  15. Successful implementation of ageing management exemplified at the cooling tower of the Emsland nuclear power plant; Erfolgreiche Umsetzung von Alterungsmanagement am Beispiel Kuehlturm des Kernkraftwerks Emsland

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Alexander [Hochtief Solutions AG, Consult IKS Energy, Frankfurt am Main (Germany). Design Kraftwerke; Dueweling, Carsten [Kernkraftwerke Lippe-Ems GmbH, Lingen (Germany). Abschnitt Bautechnik

    2013-09-01

    The paper describes the successful implementation of the restoration of water distribution channels at the cooling tower of the Emsland nuclear power plant under the aspect of ageing management. The main challenge of ageing management is the determination of potential ageing mechanism and to avoid systematically and effectively their damaging influences. In the course of the annual site inspections, abnormalities at the lower side of the water-distribution channels of the cooling tower were detected, analysed, and repaired. The procedures conserve the load bearing reinforcement only for a certain period. Therefore permanent structural monitoring is needed. (orig.)

  16. Analysis of the accident in the second power-generating unit of the Chernobyl nuclear power plant caused by inadequate makeup of the reactor cooling loop

    International Nuclear Information System (INIS)

    Vasil'chenko, V.N.; Kramerov, A.Ya.; Mikhailov, D.A.

    1995-01-01

    The accident in the second power-generating unit of the Chernobyl nuclear power plant on October 11, 1991 was the result of unauthorized connection of the TG-4 turbogenerator, which was shut down for repairs, into the grid (in the off-design asynchronous engine mode), and this resulted in a serious fire in the machine room and subsequent failure of systems which are important for safety and which ensure the design mode of reactor cooling: These were primarily failures of the feed and emergency feed pumps and failure of the BRU-B control valve, which regulates steam release during cooling

  17. A study on nuclear heat load tolerable for NET/TF coils cooled by internal flow of helium II

    International Nuclear Information System (INIS)

    Hofmann, A.

    1988-02-01

    NbTi cables cooled by internal flow of superfluid helium are considered an option for the design of NET/TF coils with about 11 T peak fields. Starting from an available winding cross section of 0.61x0.61 m 2 for a 8 MA turns coil made of a 16 kA conductor it is shown that sufficient hydraulic cross section can be provided within such cables to remove the expected thermal load resulting from nuclear heating with exponential decay from inboard to outboard side of the winding. The concept is a pancake type coil with 1.8 K helium fed-in the high field region of each pancake. The temperature distribution within such coils is calculated, and the local safety margin is determined from temperature and field. The calculation takes account of nuclear and a.c. heating, and of thermal conductance between the individual layers and the coil casing. It is shown that operation with 1.8 K inlet and about 3 K outlet temperature is possible. The electrical insulation with about 0.5 mm thickness proves to provide sufficient thermal insulation. No additional thermal shield is required between the coil casing and the winding package. Two different types of conductors are being considered: a) POLO type cable with quadratic cross section and a central circular coolant duct, and b) an LCT type cable with two conductors wound in hand. Both concepts with about 500 m length of the cooland channels are shown to meet the requirements resulting from a peak nuclear heat load of 0.3 mW/cm 3 in the inboard turns. The hydraulic diameters are sufficient to operate each coils with self-sustained fountain effect pumps. Even appreciably higher heat loads with up to 3 mW/cm 3 of nuclear heating can be tolerated for the POLO type cable when the hydraulic diameter is enlarged to its maximum of 17 mm. (orig.) [de

  18. Initial assessment of environmental effects on SiC/SiC composites in helium-cooled nuclear systems

    Energy Technology Data Exchange (ETDEWEB)

    Contescu, Cristian I [ORNL

    2013-09-01

    This report summarized the information available in the literature on the chemical reactivity of SiC/SiC composites and of their components in contact with the helium coolant used in HTGR, VHTR and GFR designs. In normal operation conditions, ultra-high purity helium will have chemically controlled impurities (water, oxygen, carbon dioxide, carbon monoxide, methane, hydrogen) that will create a slightly oxidizing gas environment. Little is known from direct experiments on the reactivity of third generation (nuclear grade) SiC/SiC composites in contact with low concentrations of water or oxygen in inert gas, at high temperature. However, there is ample information about the oxidation in dry and moist air of SiC/SiC composites at high temperatures. This information is reviewed first in the next chapters. The emphasis is places on the improvement in material oxidation, thermal, and mechanical properties during three stages of development of SiC fibers and at least two stages of development of the fiber/matrix interphase. The chemical stability of SiC/SiC composites in contact with oxygen or steam at temperatures that may develop in off-normal reactor conditions supports the conclusion that most advanced composites (also known as nuclear grade SiC/SiC composites) have the chemical resistance that would allow them maintain mechanical properties at temperatures up to 1200 1300 oC in the extreme conditions of an air or water ingress accident scenario. Further research is needed to assess the long-term stability of advanced SiC/SiC composites in inert gas (helium) in presence of very low concentrations (traces) of water and oxygen at the temperatures of normal operation of helium-cooled reactors. Another aspect that needs to be investigated is the effect of fast neutron irradiation on the oxidation stability of advanced SiC/SiC composites in normal operation conditions.

  19. Physical parameters of effluent from nuclear power station cooling towers; Fizicki parametri efluenata iz rashladnih tornjeva nuklearne elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Vehauc, A [Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia)

    1992-07-01

    Physical parameters of the effluent dispersed from the wet cooling towers, i.e. mixture of the warm moist air with the entrained droplets are analysed. Understanding of the effluent physical parameters at the exit of cooling tower is important for prediction of the effluent dispersion in the environment. Mass and droplet diameter distributors of the drifted cooling water are measured in situ and also, drift eliminators are characterised experimentally. A new numerical method for heat and mass transfer evaluation in the cooling tower packing (fill) was developed, that leads to more accurate prediction for outlet air parameters in relation of plant power rate, cooling tower characteristics and atmospheric conditions. (author)

  20. Influence of cooling water discharges from Kaiga Nuclear Power Plant on aquatic ecology of the Kadra reservoir

    International Nuclear Information System (INIS)

    Ghosh, T.K.; Zargar, S.; Kulkarni, A.V.

    2007-01-01

    The alterations induced in the ambient temperature can lead to wide manifestations in species distribution and community structure. In general, elevated water temperature causes changes in species composition, species dominance, standing crop and productivity of biota including phytoplankton communities in any aquatic ecosystem. Thus warm water discharges from power plants into the receiving water bodies may adversely affect aquatic ecology. In the absence of exhaustive data on the response of aquatic organisms and ecosystems in the tropics to elevated temperatures, the only option is to draw inferences, from the experiences in the subtropical and temperature areas. Since, sufficient data on similar line are not available in tropical environment, present paper delineates certain aspects of aquatic ecology of the Kadra reservoir where cooling water is discharged. The study suggests the heated effluents from Kaiga Nuclear Power plant caused changes in dissolved oxygen and pH of water, heterotrophic bacterial population, sediment biogeochemical cycles related biochemical processes, species composition, species dominance, standing crop and productivity of biota including phytoplankton communities within 500 m from End of Discharge Canal point of Kadra reservoir when two units are running in full capacity. (author)

  1. Interactive Real-time Simulation of a Nuclear Reactor Emergency Core Cooling System on a Desktop Computer

    International Nuclear Information System (INIS)

    Muncharoen, C.; Chanyotha, S.; Bereznai, G.

    1998-01-01

    The simulation of the Emergency Core Cooling System for a 900 MW nuclear power plant has been developed by using object oriented programming language. It is capable of generating code that executes in real-time on a PENTIUM 100 or equivalent personal computer. Graphical user interface ECCS screens have been developed using Lab VIEW to allow interactive control of ECCS. The usual simulator functions, such as freeze, run, iterate, have been provided, and a number of malfunctions may be activated. A large pipe break near the reactor inlet header has been simulated to verify the response of the ECCS model. LOCA detection, ECC initiation, injection and recovery phased are all modeled, and give results consistent with safety analysis data for a 100% break. With stand alone ECCS simulation, the changes of flow and pressure in ECCS can be observed. The operator can study operational procedures and get used to LOCA in case of the LOCA. Practicing with malfunction, the operator will improve problem solving skills and gain a deeper comprehension of ECCS

  2. Background and anthropogenic radionuclide derived dose rates to freshwater ecosystem - Nuclear power plant cooling pond - Reference organisms

    International Nuclear Information System (INIS)

    Nedveckaite, T.; Filistovic, V.; Marciulioniene, D.; Prokoptchuk, N.; Plukiene, R.; Gudelis, A.; Remeikis, V.; Yankovich, T.; Beresford, N.-A.

    2011-01-01

    The radiological assessment of non-human biota to demonstrate protection is now accepted by a number of international and national bodies. Therefore, it is necessary to develop a scientific basis to assess and evaluate exposure of biota to ionizing radiation. Radionuclides from the Ignalina Nuclear Power Plant (Lithuania) were discharged into Lake Druksiai cooling pond. Additional radionuclide migration and recharge to this lake from a hypothetical near-surface, low-level radioactive waste disposal, to be situated 1.5 km from the lake, had been simulated using RESRAD-OFFSITE code. This paper uses ERICA Integrated Approach with associated tools and databases to compare the radiological dose to freshwater reference organisms. Based on these data, it can be concluded that background dose rates to non-human biota in Lake Druksiai far exceed those attributable to anthropogenic radionuclides. With respect the fishery and corresponding annual committed effective human dose as a result of this fish consumption Lake Druksiai continues to be a high-productivity water body with intensive angling and possible commercial fishing. - Highlights: → Dose rates to the reference organisms are lower than expected from the background radioactivity. → Pelagic fish part of adult human annual committed effective dose would be as small as a few μSv y -1 . → With respect the fishery Lake Druksiai continues to be a high-productivity water body.

  3. Cooling water distribution system

    Science.gov (United States)

    Orr, Richard

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  4. Detailed Reaction Kinetics for CFD Modeling of Nuclear Fuel Pellet Coating for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Battaglia, Francine

    2008-01-01

    The research project was related to the Advanced Fuel Cycle Initiative and was in direct alignment with advancing knowledge in the area of Nuclear Fuel Development related to the use of TRISO fuels for high-temperature reactors. The importance of properly coating nuclear fuel pellets received a renewed interest for the safe production of nuclear power to help meet the energy requirements of the United States. High-temperature gas-cooled nuclear reactors use fuel in the form of coated uranium particles, and it is the coating process that was of importance to this project. The coating process requires four coating layers to retain radioactive fission products from escaping into the environment. The first layer consists of porous carbon and serves as a buffer layer to attenuate the fission and accommodate the fuel kernel swelling. The second (inner) layer is of pyrocarbon and provides protection from fission products and supports the third layer, which is silicon carbide. The final (outer) layer is also pyrocarbon and provides a bonding surface and protective barrier for the entire pellet. The coating procedures for the silicon carbide and the outer pyrocarbon layers require knowledge of the detailed kinetics of the reaction processes in the gas phase and at the surfaces where the particles interact with the reactor walls. The intent of this project was to acquire detailed information on the reaction kinetics for the chemical vapor deposition (CVD) of carbon and silicon carbine on uranium fuel pellets, including the location of transition state structures, evaluation of the associated activation energies, and the use of these activation energies in the prediction of reaction rate constants. After the detailed reaction kinetics were determined, the reactions were implemented and tested in a computational fluid dynamics model, MFIX. The intention was to find a reduced mechanism set to reduce the computational time for a simulation, while still providing accurate results

  5. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2000-01-01

    As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that

  6. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    International Nuclear Information System (INIS)

    Marvin, M.D.

    1978-01-01

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated

  7. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  8. Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2015-09-01

    Full Text Available The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System and LEADER (Lead-cooled European Advanced Demonstration Reactor projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs, and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

  9. Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

    International Nuclear Information System (INIS)

    Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

    2004-01-01

    The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

  10. Evaluation on Cooling Performance of Containment Fan Cooler during Design Basis Accident with Loss of Offsite Power for Kori 3 and 4 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Bok; Lee, Sang Won [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Park, Young Chan [Atomic Creative Technology Co., LTD., Daejeon (Korea, Republic of)

    2007-10-15

    The purpose of this study is to evaluate cooling performance of containment fan cooler units and to review a technical background related to Generic Letter 96-06. In case that design basis accident (DBA) and loss of offsite power (LOOP) occurs, component cooling water (CCW) pumps cannot provide the cooling water source to fan cooler units while fan coolers coast down. Fan cooler units and CCW pumps are restarted by emergency diesel generator (EDG) operation and it takes about 30 seconds. In this scenario, before the EDG restarts and CCW flowrate is restored, heated air in the containment passes through coil of fan cooler units without cooling water source. In this situation, the boiling of water in the fan cooler units may occur. Restarting of CCW pumps may bring about condensation by injected cooling water and water hammer may occur. This thermal-hydraulic effect is sensitive to system configuration, i.e system pressure, containment pressure/temperature, EDG restarting time, etc. In this study, the evaluation of containment fan cooler units was performed for Kori 3 and 4 nuclear power plant.

  11. Cooling techniques

    International Nuclear Information System (INIS)

    Moeller, S.P.

    1994-01-01

    After an introduction to the general concepts of cooling of charged particle beams, some specific cooling methods are discussed, namely stochastic, electron and laser cooling. The treatment concentrates on the physical ideas of the cooling methods and only very crude derivations of cooling times are given. At the end three other proposed cooling schemes are briefly discussed. (orig.)

  12. A Bayesian reliability study on motorized valves for the emergency core cooling, heat transport isolation and shutdown cooling systems at Gentilly-2 Nuclear Generating Station

    International Nuclear Information System (INIS)

    Smith, J.E.; Rennick, D.F.; Nainer, A.

    1996-01-01

    The objective of this is to examine operational data on 32 motorized valves in the emergency core cooling, shutdown cooling and heat transport isolation systems and determine if the evidence would support a reduction in testing frequency of these valves. The methodology used is to examine the data which has accumulated on motorized valve failures since Gentilly-2 first entered service, compare these data with similar data from other sources, and determine whether the evidence indicate that demand-based, wear out type failure mechanisms play a significant role in the recorded failures. The statistical data are then updated, using a Bayesian updating procedure, to obtain revised time based failure rates and demand based probabilities of failure on demand for the motorized valves. The revised failure rates and probabilities are then applied to the fault tree models for the systems of interest to determine what effects there would be, with the current test intervals and with extended test intervals, on the probability of failure of the systems. (author)

  13. Condition monitoring and life assessment of lake water cooled admiralty brass condenser tubes of a nuclear power plant

    International Nuclear Information System (INIS)

    De, P.K.; Ghosal, S.K.; Kutty, K.K.; Bhat, H.R.

    2000-01-01

    The present paper deals with the failure of condenser tubes in a nuclear power plant. The tubes were made of arsenical admiralty brass and were cooled using lake water. They were in service for over 25 years. So far about 1000 tubes have been plugged as they failed due to several reasons. In order to assess the remaining life of the existing tubes as well as to investigate the cause of recent tube failures, some of the tubes from the condenser were removed and examined in detail following several procedure. It was observed that in general, wall thickness of the tubes was reduced by 10- 15%. Maximum reduction in wall thickness took place near the water inlet ends. No denting type phenomenon was observed at the tube to tube-support plate crevice locations. At certain locations on ID surfaces of some tubes, small steps, 0.2 mm high, were noticed along the longitudinal direction of the tubes. ID surfaces of the tubes were covered with light gray coloured thin and adherent corrosion products decorated with red spots at places. EDAX analysis showed that these red spots were enriched with copper. While some pits were present on the ID surfaces, the OD surfaces were covered with shining black oxide film. Fracture surfaces of the tubes, which had lost much strength and broke while taking them out of the condenser, showed presence of cleavages with fatigue striations near the OD edges. Mechanical properties of the tubes as such had deteriorated significantly. The tubes were observed to have been degraded to a large extent due to localised corrosion on the ID surfaces and corrosion fatigue damage caused by flow induced vibration. Under the present operational conditions, the tubes are expected to perform satisfactorily for a limited period. (author)

  14. Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

    International Nuclear Information System (INIS)

    Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

    2007-01-01

    The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

  15. Study on the possibility of supercritical fluid extraction for reprocessing of spent nuclear fuel from high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Duan Wuhua; Zhu Liyang; Zhu Yongjun; Xu Jingming

    2011-01-01

    International interest in high temperature gas-cooled reactor (HTGR) has been increasing in recent years. It is important to study on reprocessing of spent nuclear fuel from HTGR for recovery of nuclear resource and reduction of nuclear waste. Treatment of UO 2 pellets for preparing fuel elements of the 10 MW high temperature gas-cooled reactor (HTR-10) using supercritical fluid extraction was investigated. UO 2 pellets are difficult to be directly dissolved and extracted with TBP-HNO 3 complex in supercritical CO 2 (SC-CO 2 ), and the extraction efficiency is only about 7% under experimental conditions. UO 2 pellets are also difficult to be converted completely into nitrate with N 2 O 4 . When UO 2 pellets break spontaneously into U 3 O 8 powders with particle size below 100 μm under O 2 flow and 600degc, the extraction efficiency of U 3 O 8 powders with TBP-HNO 3 complex in SC-CO 2 can reach more than 98%. U 3 O 8 powders are easy to be completely converted into nitrate with N 2 O 4 . The extraction efficiency of the nitrate product with TBP in SC-CO 2 can reach more than 99%. So it has a potential prospect that application of supercritical fluid extraction in reprocessing of spent nuclear fuel from HTGR. (author)

  16. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  17. Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1986-01-01

    The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail [fr

  18. Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

    International Nuclear Information System (INIS)

    Hahn, G.

    1974-01-01

    The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

  19. Corrosion by cooling gases in nuclear reactors; la corrosion par les gaz caloporteurs dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Darras, R. [Commissariat a l' energie atomique et aux energies alternatives - CEA, Centre de Saclay, Section d' etude de la corrosion par gaz et metaux liquides (France)

    1960-07-01

    This article begins with a review of the various materials which can be used and the cooling gases in which they may be heated, emphasis being placed on the importance of reaching temperatures as high as possible. This is followed by a few general remarks on the dry oxidation of metals and alloys, particularly with regard to diffusion phenomena and their various possible mechanisms, and also the methods of investigation employed. Finally, the behaviour of the chief nuclear materials heated in the various gases is studied successively. Materials used for fuel (metallic uranium, uranium oxide, carbides and silicides), canning materials (magnesium, aluminium, zirconium, beryllium, stainless and refractory steels), structural materials (ordinary or slightly alloyed steels), and finally moderators (graphite, beryllium oxide) are deal with in this way. This account is backed up both by the results obtained at the CEA and by work published outside or abroad up to the present day. In conclusion, every effort has been made to direct future research on the basis of the foregoing. Reprint of a paper published in Industries Atomiques - no. 9/10, 1959, p. 3-23 [French] Dans cet article, on passe tout d'abord en revue les divers materiaux utilisables et les gaz de refroidissement dans lesquels ils peuvent etre chauffes, en insistant sur l'interet d'atteindre des temperatures aussi elevees que possible. On rappelle ensuite quelques generalites sur l'oxydation seche des metaux et alliages, notamment en ce qui concerne les phenomenes de diffusion et leurs divers mecanismes possibles ainsi que les methodes d'etude. Enfin, le comportement des principaux materiaux nucleaires chauffes dans les divers gaz est etudie successivement. On traita ainsi des materiaux combustibles (uranium metallique, oxyde, carbures et siliciures d'uranium), des materiaux de gainage (magnesium, aluminium, zirconium, beryllium, aciers inoxydables et refractaires), des materiaux de structure (aciers ordinaires

  20. Semioptimal practicable algorithmic cooling

    International Nuclear Information System (INIS)

    Elias, Yuval; Mor, Tal; Weinstein, Yossi

    2011-01-01

    Algorithmic cooling (AC) of spins applies entropy manipulation algorithms in open spin systems in order to cool spins far beyond Shannon's entropy bound. Algorithmic cooling of nuclear spins was demonstrated experimentally and may contribute to nuclear magnetic resonance spectroscopy. Several cooling algorithms were suggested in recent years, including practicable algorithmic cooling (PAC) and exhaustive AC. Practicable algorithms have simple implementations, yet their level of cooling is far from optimal; exhaustive algorithms, on the other hand, cool much better, and some even reach (asymptotically) an optimal level of cooling, but they are not practicable. We introduce here semioptimal practicable AC (SOPAC), wherein a few cycles (typically two to six) are performed at each recursive level. Two classes of SOPAC algorithms are proposed and analyzed. Both attain cooling levels significantly better than PAC and are much more efficient than the exhaustive algorithms. These algorithms are shown to bridge the gap between PAC and exhaustive AC. In addition, we calculated the number of spins required by SOPAC in order to purify qubits for quantum computation. As few as 12 and 7 spins are required (in an ideal scenario) to yield a mildly pure spin (60% polarized) from initial polarizations of 1% and 10%, respectively. In the latter case, about five more spins are sufficient to produce a highly pure spin (99.99% polarized), which could be relevant for fault-tolerant quantum computing.

  1. Conduction cooling: multicrate fastbus hardware

    International Nuclear Information System (INIS)

    Makowiecki, D.; Sims, W.; Larsen, R.

    1980-11-01

    Described is a new and novel approach for cooling nuclear instrumentation modules via heat conduction. The simplicity of liquid cooled crates and ease of thermal management with conduction cooled modules are described. While this system was developed primarily for the higher power levels expected with Fastbus electronics, it has many general applications

  2. Water cooling system for sintering furnaces of nuclear fuel pellets; Sistema de enfriamiento con agua para hornos de sinterizado de pastillas de combustible nuclear

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author).

  3. Emergency reactor cooling device

    International Nuclear Information System (INIS)

    Arakawa, Ken.

    1993-01-01

    An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

  4. Regulatory Guide 1.131: Qualification tests of electric cables, field splices, and connections for light-water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Criterion III, ''Design Control,'' of Appendix B, ''Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plant,'' to 10 CFR Part 50, ''Licensing of Production and Utilization Facilities,'' requires that, where a test program is used to verify the adequacy of a specific design feature, it include suitable qualification testing of a prototype unit under the most adverse design conditions. This regulatory guide describes a method acceptable to the NRC staff for complying with the Commission's regulations with regard to qualification testing of electric cables, field splices, and connections for service in light-water-cooled nuclear power plants to ensure that the cables, field splices, and connections can perform their safety-related functions. The fire test provisions of this guide do not apply to qualification for an installed configuration

  5. Anthropogenic radionuclide fluxes and distribution in bottom sediments of the cooling basin of the Ignalina Nuclear Power Plant

    International Nuclear Information System (INIS)

    Marčiulionienė, D.; Mažeika, J.; Lukšienė, B.; Jefanova, O.; Mikalauskienė, R.; Paškauskas, R.

    2015-01-01

    Based on γ-ray emitting artificial radionuclide spectrometric measurements, an assessment of areal and vertical distribution of 137 Cs, 60 Co and 54 Mn activity concentrations in bottom sediments of Lake Drūkšiai was performed. Samples of bottom sediments from seven monitoring stations within the cooling basin were collected in 1988–1996 and 2007–2010 (in July–August). For radionuclide areal distribution analysis, samples from the surface 0–5 cm layer were used. Multi sample cores sliced 2 cm, 3 cm or 5 cm thick were used to study the vertical distribution of radionuclides. The lowest 137 Cs activity concentrations were obtained for two stations that were situated close to channels with radionuclide discharges, but with sediments that had a significantly smaller fraction of organic matter related to finest particles and consequently smaller radionuclide retention potential. The 137 Cs activity concentration was distributed quite evenly in the bottom sediments from other investigated monitoring stations. The highest 137 Cs activity concentrations in the bottom sediments of Lake Drūkšiai were measured in the period of 1988–1989; in 1990, the 137 Cs activity concentrations slightly decreased and they varied insignificantly over the investigation period. The obtained 238 Pu/ 239,240 Pu activity ratio values in the bottom sediments of Lake Drūkšiai represented radioactive pollution with plutonium from nuclear weapon tests. Higher 60 Co and 54 Mn activity concentrations were observed in the monitoring stations that were close to the impact zones of the technical water outlet channel and industrial rain drainage system channel. 60 Co and 54 Mn activity concentrations in the bottom sediments of Lake Drūkšiai significantly decreased when operations at both INPP reactor units were stopped. The vertical distribution of radionuclides in bottom sediments revealed complicated sedimentation features, which may have been affected by a number of natural and

  6. Rapid quantification of viable Legionella in nuclear cooling tower waters using filter cultivation, fluorescent in situ hybridization and solid-phase cytometry.

    Science.gov (United States)

    Baudart, J; Guillaume, C; Mercier, A; Lebaron, P; Binet, M

    2015-05-01

    To develop a rapid and sensitive method to quantify viable Legionella spp. in cooling tower water samples. A rapid, culture-based method capable of quantifying as few as 600 Legionella microcolonies per litre within 2 days in industrial waters was developed. The method combines a short cultivation step of microcolonies on GVPC agar plate, specific detection of Legionella cells by a fluorescent in situ hybridization (FISH) approach, and a sensitive enumeration using a solid-phase cytometer. Following optimization of the cultivation conditions, the qualitative and quantitative performance of the method was assessed and the method was applied to 262 nuclear power plant cooling water samples. The performance of this method was in accordance with the culture method (NF-T 90-431) for Legionella enumeration. The rapid detection of viable Legionella in water is a major concern to the effective monitoring of this pathogenic bacterium in the main water sources involved in the transmission of legionellosis infection (Legionnaires' disease). The new method proposed here appears to be a robust, efficient and innovative means for rapidly quantifying cultivable Legionella in cooling tower water samples within 48 h. © 2015 The Society for Applied Microbiology.

  7. Successful implementation of ageing management exemplified at the cooling tower of Emsland nuclear power plant; Erfolgreiche Umsetzung von Alterungsmanagement am Beispiel Kuehlturm des Kernkraftwerkes Emsland

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Alexander [Hochtief Solutions AG, Consult IKS Energy, Frankfurt am Main (Germany). Design Kraftwerke; Dueweling, Carsten [Kernkraftwerke Lippe-Ems GmbH, Lingen (Germany). Abschnitt Bautechnik

    2013-07-15

    The present paper describes the successful implementation of the restoration of water-distribution channels at the cooling tower of the Emsland nuclear power plant under the aspect of ageing management. The main challenge of aging management is the determination of potential aging mechanism and to avoid systematically and effectively their damaging influences. In the course of the annual site inspections abnormalities at the lower side of the water-distribution channels of the cooling tower were detected, analysed, and repaired. The extraordinary high chlorine equivalent of the cooling water was identified as main reason of the damages located. Due to extensive infiltration into the concrete structure, chloride-induced corrosion generates a volume expansion of the reinforcement and thereby to a blast off of the concrete covering. According to the restoration concept, the damaged concrete was removed by maximum pressure water jet blasting; where necessary the reinforcement was retrofitted and a layered concrete substitution was applied by synthetic cement mortar. The realised procedures conserve the load bearing reinforcement only for a certain period, because the permanent chloride infiltration could not be stopped. Therefore, the structure has to be monitored permanently. (orig.)

  8. Radiation Dose Assessment For The Biota Of Terrestrial Ecosystems In The Shoreline Zone Of The Chernobyl Nuclear Power Plant Cooling Pond

    International Nuclear Information System (INIS)

    Farfan, E.; Jannik, T.

    2011-01-01

    Radiation exposure of the biota in the shoreline area of the Chernobyl Nuclear Power Plant Cooling Pond was assessed to evaluate radiological consequences from the decommissioning of the Cooling Pond. The article addresses studies of radioactive contamination of the terrestrial faunal complex and radionuclide concentration ratios in bodies of small birds, small mammals, amphibians, and reptiles living in the area. The data were used to calculate doses to biota using the ERICA Tool software. Doses from 90 Sr and 137 Cs were calculated using the default parameters of the ERICA Tool and were shown to be consistent with biota doses calculated from the field data. However, the ERICA dose calculations for plutonium isotopes were much higher (2-5 times for small mammals and 10-14 times for birds) than the doses calculated using the experimental data. Currently, the total doses for the terrestrial biota do not exceed maximum recommended levels. However, if the Cooling Pond is allowed to drawdown naturally and the contaminants of the bottom sediments are exposed and enter the biological cycle, the calculated doses to biota may exceed the maximum recommended values. The study is important in establishing the current exposure conditions such that a baseline exists from which changes can be documented following the lowering of the reservoir water. Additionally, the study provided useful radioecological data on biota concentration ratios for some species that are poorly represented in the literature.

  9. ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

    International Nuclear Information System (INIS)

    Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

    2011-01-01

    During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

  10. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  11. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  12. Invariant methods for an ensemble-based sensitivity analysis of a passive containment cooling system of an AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Nicola, Giancarlo; Borgonovo, Emanuele; Zio, Enrico

    2016-01-01

    Sensitivity Analysis (SA) is performed to gain fundamental insights on a system behavior that is usually reproduced by a model and to identify the most relevant input variables whose variations affect the system model functional response. For the reliability analysis of passive safety systems of Nuclear Power Plants (NPPs), models are Best Estimate (BE) Thermal Hydraulic (TH) codes, that predict the system functional response in normal and accidental conditions and, in this paper, an ensemble of three alternative invariant SA methods is innovatively set up for a SA on the TH code input variables. The ensemble aggregates the input variables raking orders provided by Pearson correlation ratio, Delta method and Beta method. The capability of the ensemble is shown on a BE–TH code of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized water reactor AP1000, during a Loss Of Coolant Accident (LOCA), whose output probability density function (pdf) is approximated by a Finite Mixture Model (FMM), on the basis of a limited number of simulations. - Highlights: • We perform the reliability analysis of a passive safety system of Nuclear Power Plant (NPP). • We use a Thermal Hydraulic (TH) code for predicting the NPP response to accidents. • We propose an ensemble of Invariant Methods for the sensitivity analysis of the TH code • The ensemble aggregates the rankings of Pearson correlation, Delta and Beta methods. • The approach is tested on a Passive Containment Cooling System of an AP1000 NPP.

  13. Technologies for improving the availability and reliability of current and future water cooled nuclear power plants. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-11-01

    One of the activities of the IAEA is to provide all Member States with an international source of balanced, objective information on advanced in technology for water cooled reactors. Since the global nuclear industry has a common interest in improving plant availability and reliability to assure specific individual plant and country perspective as well as to have an image of well managed competitive industry, the IAEA held a Technical Committee Meeting on Technologies for Improving the Availability and Reliability of Current and Future Water Cooled Nuclear Power Plants in September 1997. The basic aim to was to identify, review and exchange information on international developments in technologies for achieving high availability and reliability and to suggest areas where further technical advances could contribute to improvement of performance. Designs for future plants were presented in the context of how they can accommodate both the organizational and technical means for reaching even higher levels of performance. This proceedings contains the contributed papers presented at this Meeting each with a separate abstract. Four sessions were concerned with: policies, practices and procedures for achieving high reliability and availability; improving availability and reliability through better use of today`s technologies; recent advances in technologies for improving availability and reliability; achieving high availability for new plants Refs, figs, tabs

  14. Technologies for improving the availability and reliability of current and future water cooled nuclear power plants. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-11-01

    One of the activities of the IAEA is to provide all Member States with an international source of balanced, objective information on advanced in technology for water cooled reactors. Since the global nuclear industry has a common interest in improving plant availability and reliability to assure specific individual plant and country perspective as well as to have an image of well managed competitive industry, the IAEA held a Technical Committee Meeting on Technologies for Improving the Availability and Reliability of Current and Future Water Cooled Nuclear Power Plants in September 1997. The basic aim to was to identify, review and exchange information on international developments in technologies for achieving high availability and reliability and to suggest areas where further technical advances could contribute to improvement of performance. Designs for future plants were presented in the context of how they can accommodate both the organizational and technical means for reaching even higher levels of performance. This proceedings contains the contributed papers presented at this Meeting each with a separate abstract. Four sessions were concerned with: policies, practices and procedures for achieving high reliability and availability; improving availability and reliability through better use of today's technologies; recent advances in technologies for improving availability and reliability; achieving high availability for new plants

  15. A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2011-07-01

    This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)

  16. Lunar nuclear power plant design for thermal-hydraulic cooling in nano-scale environment: Nuclear engineering-based interdisciplinary nanotechnology

    International Nuclear Information System (INIS)

    Woo, Tae Ho

    2015-01-01

    The environment of the Moon is nearly vacant, which has very low density of several kinds of gases. It has the molecular level contents of the lunar atmosphere in Table 1, which is recognized that radiation heat transfer is a major cooling method. The coolant of the nuclear power plant (NPP) in the lunar base is the Moon surface soil , which is known as the regolith. The regolith is the layer of loose and heterogeneous material covering the solid rock. For finding the optimized length of the radiator of the coolant in the lunar NPP, the produced power and Moon environmental temperature are needed. This makes the particular heat transfer characteristics in heat transfer in the Moon surface. The radiation is the only heat transfer way due to very weak atmosphere. It is very cold in the night time and very hot in the daytime on the surface of the ground. There are comparisons between lunar high land soil and Earth averages in Table 2. In the historical consideration, Konstantin Tsiolkovsky made a suggestion for the colony on the Moon.. There are a number of ideas for the conceptual design which have been proposed by several scientists. In 1954, Arthur C. Clarke mentioned a lunar base of inflatable modules covered in lunar dust for insulation. John S. Rinehart suggested the structure of the stationary ocean of dust, because there could be a mile-deep dust ocean on the Moon, which gives a safer design. In 1959, the project horizon was launched regarding the U.S. Army's plan to establish a fort on the Moon by 1967. H. H. Koelle, a German rocket engineer of the Army Ballistic Missile Agency, leaded the project (ABMA). There was the first landing in 1965 and 245 tons of cargos were transported to the outpost by 1966. The coolant material of regolith in the Moon is optimized for the NPP. By the simulation, there are some results. The temperature is calculated as the 9 nodals by radiation heat transfer from the potassium coolant to the regolith flow. The high efficiency

  17. Lunar nuclear power plant design for thermal-hydraulic cooling in nano-scale environment: Nuclear engineering-based interdisciplinary nanotechnology

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Systemix Global Co. Ltd., Seoul (Korea, Republic of)

    2015-05-15

    The environment of the Moon is nearly vacant, which has very low density of several kinds of gases. It has the molecular level contents of the lunar atmosphere in Table 1, which is recognized that radiation heat transfer is a major cooling method. The coolant of the nuclear power plant (NPP) in the lunar base is the Moon surface soil , which is known as the regolith. The regolith is the layer of loose and heterogeneous material covering the solid rock. For finding the optimized length of the radiator of the coolant in the lunar NPP, the produced power and Moon environmental temperature are needed. This makes the particular heat transfer characteristics in heat transfer in the Moon surface. The radiation is the only heat transfer way due to very weak atmosphere. It is very cold in the night time and very hot in the daytime on the surface of the ground. There are comparisons between lunar high land soil and Earth averages in Table 2. In the historical consideration, Konstantin Tsiolkovsky made a suggestion for the colony on the Moon.. There are a number of ideas for the conceptual design which have been proposed by several scientists. In 1954, Arthur C. Clarke mentioned a lunar base of inflatable modules covered in lunar dust for insulation. John S. Rinehart suggested the structure of the stationary ocean of dust, because there could be a mile-deep dust ocean on the Moon, which gives a safer design. In 1959, the project horizon was launched regarding the U.S. Army's plan to establish a fort on the Moon by 1967. H. H. Koelle, a German rocket engineer of the Army Ballistic Missile Agency, leaded the project (ABMA). There was the first landing in 1965 and 245 tons of cargos were transported to the outpost by 1966. The coolant material of regolith in the Moon is optimized for the NPP. By the simulation, there are some results. The temperature is calculated as the 9 nodals by radiation heat transfer from the potassium coolant to the regolith flow. The high efficiency

  18. Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

    International Nuclear Information System (INIS)

    Anfimov, S.S.

    2001-01-01

    In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future

  19. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  20. Device for regulating light water nuclear reactors by changing the boric acid concentration in the cooling water circuit

    International Nuclear Information System (INIS)

    Brown, W.W.; Van der Schoot, M.R.

    1980-01-01

    Small changes in boric acid concentration can be carried out quickly by a combination of an ion exchanger with temperature-dependent capacity and an evaporator. No boric acid need be extracted from the circuit or added to it. However, if large changes of concentration are required, boric acid has to be added. The evaporator is then used to separate distilled water and concentrated boric acid when the cooling water is diluted. (DG) [de

  1. The largest Fresco in Europe on cooling tower of nuclear power station of Cruas Meysse in Ardeche, France

    International Nuclear Information System (INIS)

    Di Mayo, J.L.

    1993-01-01

    The Nuclear Power Station Cruas Meysse is on the most important communication way of France, in the Rhone Valley, between the Rhin and the Mediterranean Sea. In the South of the Rhone Valley, the Nuclear Power Plant is situated near the very important site of 'Tricastin', the largest nuclear area in France. Cruas Meysse has a very good integration to the economy, social, and cultural scheme ; that's why EDF and the Ardeche Department had enter into partnership to associate art and technology of our time, and offer a work for everybody - 'Le Verseau' is the largest fresco in Europe - It gives a gigantic signalling system to the Ardeche Department, because the Nuclear Power Station has a very interesting position, close the motor way A7, the National 7 road, and the way of high speed train (TGV) an another symbol of the high French technology

  2. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... INFORMATION CONTACT: Stephen C. Burton, Region III, telephone: 301-415-7000; email: Stephen[email protected] , U... at Rockville, Maryland, this 7th day of October 2013. For the Nuclear Regulatory Commission. Thomas H...

  3. Proposal for the small high temperature gas cooled reactor (application of nuclear energy in the changing and emerging energy markets)

    International Nuclear Information System (INIS)

    Remond R Pahladsingh

    2005-01-01

    With the peaking of the oil-production (2006) on the global energy-markets, already predicted by M.K. Hubbert in 1970 and the limited possibilities to use coal as an alternative, the world has come in a critical situation. Exponential growth in global population and the demand of energy (IEA) have created a unique, but known, problem in the present time. The difference with the past is the period in which the changes will take place is not centuries or decades but most likely a few years. There is no alternative for oil to anticipate on the coming energy-crisis. The coal-industry has been demolished in Europe and America, under the pressure of the environmental lobby, and the nuclear industry is in very bad shape. Large scale nuclear plants can only be built in strong power grids; the present power grids are exposed to de-investment and less maintenance in liberalized and de-regulated economies. Furthermore the Digital Society is placing additional requirements on reliability and quality of our electric power systems. Life extension programs in the nuclear industry have shown that the present industry is robust, safety standards are high but at the same time no breakthrough technology has come to the market. The manpower in the nuclear industry is ageing. There are no real indications of growth in the large scale nuclear industry. The present industry is heavily relaying on the present grid concepts that have shown many difficulties in recent past and the increased failures in power-grids will create constraints to excel the option of nuclear energy in the electric power industry This paper is a request to the nuclear industry to come with a revival program, easy to implement nuclear technology and to help the world to solve the huge energy-problems that present and future generations are facing in the world. The proposal in the paper is a request to bring nuclear energy back in our society (Yes in my backyard-YIMBY) so that the many fruitful uses of nuclear energy

  4. Cooling towers

    International Nuclear Information System (INIS)

    Boernke, F.

    1975-01-01

    The need for the use of cooling systems in power plant engineering is dealt with from the point of view of a non-polluting form of energy production. The various cooling system concepts up to the modern natural-draught cooling towers are illustrated by examples. (TK/AK) [de

  5. Cooling towers for thermal power plants

    International Nuclear Information System (INIS)

    Chaboseau, J.

    1987-01-01

    After a brief recall on cooling towers testing and construction, this paper presents four examples of very large French nuclear power plant cooling towers, and one of an Australian thermal power plant [fr

  6. Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

    International Nuclear Information System (INIS)

    Langton, S.E.; Buijs, A.; Pencer, J.

    2015-01-01

    Highlights: • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling sensitivity. • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling uncertainty. • Isotopes of Zr make large contributions to SCWR modelling uncertainty. - Abstract: Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor k eff are 6.31 mk for the SCWR at the beginning of cycle (BOC) and 6.99 mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02 mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0 mk and 0.9 mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1 mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium

  7. Sensitivity study of a method for updating a finite element model of a nuclear power station cooling tower

    International Nuclear Information System (INIS)

    Billet, L.

    1994-01-01

    The Research and Development Division of Electricite de France is developing a surveillance method of cooling towers involving on-site wind-induced measurements. The method is supposed to detect structural damage in the tower. The damage is identified by tuning a finite element model of the tower on experimental mode shapes and eigenfrequencies. The sensitivity of the method was evaluated through numerical tests. First, the dynamic response of a damaged tower was simulated by varying the stiffness of some area of the model shell (from 1 % to 24 % of the total shell area). Second, the structural parameters of the undamaged cooling tower model were updated in order to make the output of the undamaged model as close as possible to the synthetic experimental data. The updating method, based on the minimization of the differences between experimental modal energies and modal energies calculated by the model, did not detect a stiffness change over less than 3 % of the shell area. Such a sensitivity is thought to be insufficient to detect tower cracks which behave like highly localized defaults. (author). 8 refs., 9 figs., 6 tabs

  8. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  9. Computation, measurement and analysis of the reactivity-to-power-transfer-function for the sodium cooled nuclear power plant KNK I

    International Nuclear Information System (INIS)

    Hoppe, P.; Mitzel, F.

    1977-02-01

    The Reactivity-to-Power-Transfer-Function for the sodium cooled nuclear power plant KNK I (Kompakte Natriumgekuehlte Kernenergieanlage) has been measured and compared with theoretical results. The measurements have been performed with the help of pseudostochastic reactivity perturbations. The transfer function has been determined by computing the auto- and cross-power-spectral-densities for the reactivity- and neutron flux signals. The agreement between the experimental and theoretical transfer function could be improved by adjusting the reactivity coefficients. The applications of these measurements with respect to reactor diagnosis and malfunction detection are discussed. For this purpose the accuracy of the measured transfer function is of great importance. Therefore an extensive error analysis has been performed. It turned out, that the inherent instability of the reactor without control system and the feedback by the primary coolant system were the reasons for comparatively big systematical errors. The conditions have been derived under which these types of errors can be considerably reduced. The conclusions can also be applied to analogical measurements at fast sodium cooled reactors. Because of their inherent stability the systematical errors will be reduced. (orig.) [de

  10. Experimental facility with two-phase flow and with high concentration of non-condensable gases for research and development of emergency cooling system of advanced nuclear reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

    2006-01-01

    The development of emergency cooling passive systems of advanced nuclear reactors requires the research of some relative processes to natural circulation, in two-phase flow conditions involving condensation processes in the presence of non-condensable gases. This work describes the main characteristics of the experimental facility called Bancada de Circulacao Natural (BCN), designed for natural circulation experiments in a system with a hot source, electric heater, a cold source, heat exchanger, operating with two-phase flow and with high concentration of noncondensable gas, air. The operational tests, the data acquisition system and the first experimental results in natural circulation are presented. The experiments are transitory in natural circulation considering power steps. The distribution of temperatures and the behavior of the flow and of the pressure are analyzed. The experimental facility, the instrumentation and the data acquisition system demonstrated to be adapted for the purposes of research of emergency cooling passive systems, operating with two-phase flow and with high concentration of noncondensable gases. (author)

  11. BLACK HOLE-NEUTRON STAR MERGERS WITH A HOT NUCLEAR EQUATION OF STATE: OUTFLOW AND NEUTRINO-COOLED DISK FOR A LOW-MASS, HIGH-SPIN CASE

    International Nuclear Information System (INIS)

    Deaton, M. Brett; Duez, Matthew D.; Foucart, Francois; O'Connor, Evan; Ott, Christian D.; Scheel, Mark A.; Szilagyi, Bela; Kidder, Lawrence E.; Muhlberger, Curran D.

    2013-01-01

    Neutrino emission significantly affects the evolution of the accretion tori formed in black hole-neutron star mergers. It removes energy from the disk, alters its composition, and provides a potential power source for a gamma-ray burst. To study these effects, simulations in general relativity with a hot microphysical equation of state (EOS) and neutrino feedback are needed. We present the first such simulation, using a neutrino leakage scheme for cooling to capture the most essential effects and considering a moderate mass (1.4 M ☉ neutron star, 5.6 M ☉ black hole), high-spin (black hole J/M 2 = 0.9) system with the K 0 = 220 MeV Lattimer-Swesty EOS. We find that about 0.08 M ☉ of nuclear matter is ejected from the system, while another 0.3 M ☉ forms a hot, compact accretion disk. The primary effects of the escaping neutrinos are (1) to make the disk much denser and more compact, (2) to cause the average electron fraction Y e of the disk to rise to about 0.2 and then gradually decrease again, and (3) to gradually cool the disk. The disk is initially hot (T ∼ 6 MeV) and luminous in neutrinos (L ν ∼ 10 54 erg s –1 ), but the neutrino luminosity decreases by an order of magnitude over 50 ms of post-merger evolution

  12. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing

  13. Evaluation of fuel fabrication and the back end of the fuel cycle for light-water- and heavy-water-cooled nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carter, W.L.; Olsen, A.R.

    1979-06-01

    The classification of water-cooled nuclear reactors offers a number of fuel cycles that present inherently low risk of weapons proliferation while making power available to the international community. Eight fuel cycles in light water reactor (LWR), heavy water reactor (HWR), and the spectral shift controlled reactor (SSCR) systems have been proposed to promote these objectives in the International Fuel Cycle Evaluation (INFCE) program. Each was examined in an effort to provide technical and economic data to INFCE on fuel fabrication, refabrication, and reprocessing for an initial comparison of alternate cycles. The fuel cycles include three once-through cycles that require only fresh fuel fabrication, shipping, and spent fuel storage; four cycles that utilize denatured uranium--thorium and require all recycle operations; and one cycle that considers the LWR--HWR tandem operation requiring refabrication but no reprocessing.

  14. Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2001-01-01

    In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

  15. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 1

    International Nuclear Information System (INIS)

    Price, M.S.T.; Lafontaine, I.

    1985-01-01

    The purpose of this volume is to introduce the main types of nuclear reactor in the European Community (EC), select reference plants for further study, estimate the waste streams from the reference reactors, survey the transport regulations and assess existing containers

  16. Possibilities for the reduction of the activity build-up in the primary circuit of water-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Sachse, G.; Mittag, I.

    1985-01-01

    Basing upon the international literature in a review are refered: research and development efforts; release, transport and deposition of radioactive corrosion products under primary circuit conditions; experimental results in test and technical systems; and possibilities to control radiation fields in nuclear power plants by waterchemical measures, decontamination, and high temperature filtration. Relevant patents are summarized. (author)

  17. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-15

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR.

  18. Assessment of Proliferation Resistance of Closed Nuclear Fuel Cycle System with Sodium Cooled Fast Reactors Using INPRO Evaluation Methodology

    International Nuclear Information System (INIS)

    Kim, Young In; Hahn, Do Hee; Won, Byung Chool; Lee, Dong Uk

    2007-11-01

    Using the INPRO methodology, the proliferation resistance of an innovative nuclear energy system(INS) defined as a closed nuclear fuel cycle system consisting of KALIMER and pyroprocessing, has been assessed. Considering a very early development stage of the INS concept, the PR assessment is carried out based on intrinsic features, if required information and data are not available. The PR assessment of KALIMER and JSFR using the INPRO methodology affirmed that an adequate proliferation resistance has been achieved in both INSs CNFC-SFR, considering the assessor's progress and maturity of design development. KALIMER and JSFR are developed or being developed conforming to the targets and criteria defined for developing Gen IV nuclear reactor system. Based on these assessment results, proliferation resistance and physical protection(PR and PP) of KALIMER and JSFR are evaluated from the viewpoint of requirements for future nuclear fuel cycle system. The envisioned INSs CNFC-SFR rely on active plutonium management based on a closed fuel cycle, in which a fissile material is recycled in an integrated fuel cycle facility within proper safeguards. There is no isolated plutonium in the closed fuel cycle. The material remains continuously in a sequence of highly radioactive matrices within inaccessible facilities. The proliferation resistance assessment should be an ongoing analysis that keeps up with the progress and maturity of the design of Gen IV SFR

  19. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  20. Experience in the development of metal uranium-base nuclear fuel for heavy-water gas-cooled reactors

    International Nuclear Information System (INIS)

    Ashikhmin, V.P.; Vorob'ev, M.A.; Gusarov, M.S.; Davidenko, A.S.; Zelenskij, V.F.; Ivanov, V.E.; Krasnorutskij, V.S.; Petel'guzov, I.A.; Stukalov, A.I.

    1978-01-01

    Investigations were carried out to solve the problem of making the development of radiation-resistant uranium fuel for power reactors including the heavy-water gas-cooled KS-150 reactor. Factors are considered that limit the lifetime of uranium fuel elements, and the ways of suppressing them are discussed. Possible reasons of the insufficient radiation resistance of uranium rod fuel element and the progress attained are analyzed. Some general problems on the fuel manufacture processes are discussed. The main results are presented on the operation of the developed fuel in research reactor loops and the commercial heavy-water KS-150 reactor. The results confirm an exceptionally high radiation resistance of fuel to burn-ups of 1.5-2%. The successful solution of a large number of problems associated with the development of metal uranium fuel provides for new possibilities of using metal uranium in power reactors

  1. Spent nuclear fuel project cold vacuum drying facility tempered water and tempered water cooling system design description

    International Nuclear Information System (INIS)

    IRWIN, J.J.

    1998-01-01

    This document provides the System Design Description (SDD) for the Cold Vacuum Drying Facility (CVDF) Tempered Water (TW) and Tempered Water Cooling (TWC) System . The SDD was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of Processing Systems (Garvin 1998), The HNF-SD-SNF-DRD-O02, 1998, Cold Vacuum Drying Facility Design Requirements, and the CVDF Design Summary Report. The SDD contains general descriptions of the TW and TWC equipment, the system functions, requirements and interfaces. The SDD provides references for design and fabrication details, operation sequences and maintenance. This SOD has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved

  2. System code assessment with thermal-hydraulic experiment to develop helium cooled breeding blanket for nuclear fusion reactor

    International Nuclear Information System (INIS)

    Yum, S. B.; Park, I. W.; Park, G. C.; Lee, D. W.

    2012-01-01

    By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He Cooled Molten Lithium (HCML) Test Blanket Module (TBM) for testing in the International Thermonuclear Experimental Reactor (ITER). A performance analysis for the thermal-hydraulics and a safety analysis for an accident caused by a loss of coolant for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs Multicomponent Mixture Analysis), which was developed by the Gas Cooled Reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM First Wall (FW) mock-up made from the same material as tho KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 11, 19, and 29 bar, and under various ranges of flow rate from 0.63 to 2.44kg/min with a constant wall temperature condition. In the present study, a thermal-hydraulic test was performed with the newly constructed helium supplying system, In which the design pressure and temperature were 9 MPa and 500 .deg. C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 8 MPa pressure, 0.2 kg/s flow rate, which are almost same conditions of the KO TBM FW. One-side of the mock-up was heated with a constant heat flux of 0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea Heat Load Test Facility-2). The wall temperatures were measured using installed thermocouples, and they show a strong parity with the code results simulated under the same test conditions

  3. System of large transport containers for waste from dismantling light water and gas-cooled nuclear reactors. Volume 2

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1986-09-01

    This report deals with the operational, radiological and economic aspects of transport as well as conceptual designs of large containers for the transport of radioactive decommissioning wastes from nuclear power plants within the member states of the European Economic Community. The means of transport, the costs and radiological detriment are considered, and conceptual designs of containers are described. Recommendations are made for further studies. (U.K.)

  4. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO{sub 2} emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO{sub 2} emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus{sup TM} Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition.

  5. Modeling and Simulation of the Sulfur-Iodine Process Coupled to a Very High-Temperature Gas-Cooled Nuclear Reactor

    International Nuclear Information System (INIS)

    Shin, Youngjoon; Lee, Taehoon; Lee, Kiyoung; Kim, Minhwan

    2015-01-01

    Hydrogen produced from water using nuclear energy will avoid both the use of fossil fuel and CO 2 emission presumed to be the dominant reason for global warming. A thermo-chemical sulfur-iodine (SI) process coupled to a Very High Temperature Gas-Cooled Reactor(VHTR) is one of the most prospective hydrogen production methods that split water using nuclear energy because the SI process is suitable for large-scale hydrogen production without CO 2 emission. The dynamic simulation code to evaluate the start-up behavior of the chemical reactors placed on the secondary helium loop of the SI process has been developed and partially verified using the steady state values obtained from the Aspen Plus TM Code simulation. As the start-up dynamic simulation results of the SI process coupled to the IHX, which is one of components in the VHTR system, it is expected that the integrated secondary helium loop of the SI process can be successfully and safely approach the steady state condition

  6. Methodology to monitor and diagnostic vibrations of the motor-pumps used in the primary cooling system of IEAR-1 nuclear research reactor

    International Nuclear Information System (INIS)

    Benevenuti, Erion de Lima

    2004-01-01

    The objectives of this study are to establish a strategy to monitor and diagnose vibrations of the motor pumps used in the primary reactor cooling system of the IEA-R1 nuclear research reactor, to verify the possibility of using the existing installed monitoring vibration system and to implement such strategy in a continuous way. Four types of mechanical problems were considered: unbalancing, misalignment, gaps and faults in bearings. An adequate set of analysis tools, well established by the industry, was selected. These are: global measurements of vibration, velocity spectrum and acceleration envelope spectrum. Three sources of data and information were used; the data measured from the primary pumps, experimental results obtained with a Spectra Quest machine used to simulate mechanical defects and data from the literature. The results show that, for the specific case of the motor-pumps of IEA-R1 nuclear research reactor, although the technique using the envelope of acceleration, which is not available in the current system used to monitor the vibration of the motor pumps, is the one with best performance, the other techniques available in the system are sufficient to monitor the four types of mechanical problems mentioned. The proposed strategy is shown and detailed in this work. (author)

  7. Laser cooling and trapping of atoms

    International Nuclear Information System (INIS)

    Chu, S.

    1995-01-01

    The basic ideas of laser cooling and atom trapping will be discussed. These techniques have applications in spectroscopy, metrology, nuclear physics, biophysics, geophysics, and polymer science. (author)

  8. Spray cooling

    International Nuclear Information System (INIS)

    Rollin, Philippe.

    1975-01-01

    Spray cooling - using water spraying in air - is surveyed as a possible system for make-up (peak clipping in open circuit) or major cooling (in closed circuit) of the cooling water of the condensers in thermal power plants. Indications are given on the experiments made in France and the systems recently developed in USA, questions relating to performance, cost and environmental effects of spray devices are then dealt with [fr

  9. 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)

  10. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  11. Corrosion study of heat exchanger tubes in pressurized water cooled nuclear reactors by conversion electron Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Homonnay, Z.; Kuzmann, E.; Varga, K.; Nemeth, Z.; Szabo, A.; Rado, K.; Schunk, J.; Tilky, P.; Patek, G.

    2005-01-01

    Nuclear energy production tends to return into the focus of interest because of the constantly increasing energy need of the world and the green house effect problems of the strongest competitor oil or gas based power plants. In addition to the construction of new nuclear power plants, lifetime extension of the existing ones is the most cost effective investment in the energy business. However, feasibility and safety issues become very important at this point, and corrosion of the construction materials should be carefully investigated before decision on a potential lifetime extension of a reactor. 57 Fe-Conversion Electron Moessbauer Spectroscopy (CEMS) is a sensitive tool to analyze the phase composition of corrosion products on the surface of stainless steel. The upper ∼300 nm can be investigated due to the penetration range of conversion electrons. The corrosion state of heat exchanger tubes from the four reactor units of the Paks Nuclear Power Plant, Hungary, were analyzed by several methods including CEMS. The primary circuit side of the tubes was studied on selected samples cut out from the heat exchangers during regular maintenance. Cr- and Ni-substituted magnetite, sometimes hematite, amorphous Fe-oxides/oxyhydroxides as well as the signal of bulk austenitic steel of the tubes were detected. The level of Cr- and Ni-substitution in the magnetite phase could be estimated from the Moessbauer spectra. Correlation between earlier decontamination cycles and the corrosion state of the heat exchangers was sought. In combination with other methods, a hybrid structure of the surface oxide layer of several microns was established. It is suggested that previous AP-CITROX decontamination cycles can be responsible for this structure which makes the oxide layer mobile. This mobility may be responsible for unwanted corrosion product transport into the reactor vessel by the primary coolant.

  12. Modern cooling systems in thermal power plants relieve environmental pollution. Pt. 2

    International Nuclear Information System (INIS)

    Brosche, D.

    1983-01-01

    Direct and indirect dry recirculation cooling, wet cooling tower, natural-draught wet cooling tower, combined cooling processes, hybrid cooling systems, cell cooling systems, auxiliary water preparation, cooling process design, afterheat removal in nuclear power plants, environmental effects, visible plumes as a function of weather conditions, environmental protection and energy supply assurance. (orig.) [de

  13. Pressure loss coefficient and flow rate of side hole in a lower end plug for dual-cooled annular nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang-Hwan, E-mail: shinch@kaeri.re.kr; Park, Ju-Yong, E-mail: juyong@kaeri.re.kr; In, Wang-Kee, E-mail: wkin@kaeri.re.kr

    2013-12-15

    Highlights: • A lower end plug with side flow holes is suggested to provide alternative flow paths of the inner channel. • The inlet loss coefficient of the lower end plug is estimated from the experiment. • The flow rate through the side holes is estimated in a complete entrance blockage of inner channel. • The consequence in the reactor core condition is evaluated with a subchannel analysis code. - Abstract: Dual-cooled annular nuclear fuel for a pressurized water reactor (PWR) has been introduced for a significant increase in reactor power. KAERI has been developing a dual-cooled annular fuel for a power uprate of 20% in an optimized PWR in Korea, the OPR1000. This annular fuel can help decrease the fuel temperature substantially relative to conventional cylindrical fuel at a power uprate. Annular fuel has dual flow channels around itself; however, the inner flow channel has a weakness in that it is isolated unlike the outer flow channel, which is open to other neighbouring outer channels for a coolant exchange in the reactor core. If the entrance of the inner channel is, as a hypothetical event, completely blocked by debris, the inner channel will then experience a rapid increase in coolant temperature such that a departure from nucleate boiling (DNB) may occur. Therefore, a remedy to avoid such a postulated accident is indispensable for the safety of annular fuel. A lower end plug with side flow holes was suggested to provide alternative flow paths in addition to the central entrance of the inner channel. In this paper, the inlet loss coefficient of the lower end plug and the flow rate through the side holes were estimated from the experimental results even in a complete entrance blockage of the inner channel. An optimization for the side hole was also performed, and the results are applied to a subchannel analysis to evaluate the consequence in the reactor core condition.

  14. BLACK HOLE-NEUTRON STAR MERGERS WITH A HOT NUCLEAR EQUATION OF STATE: OUTFLOW AND NEUTRINO-COOLED DISK FOR A LOW-MASS, HIGH-SPIN CASE

    Energy Technology Data Exchange (ETDEWEB)

    Deaton, M. Brett; Duez, Matthew D. [Department of Physics and Astronomy, Washington State University, Pullman, WA 99164 (United States); Foucart, Francois; O' Connor, Evan [Canadian Institute for Theoretical Astrophysics, University of Toronto, Toronto, Ontario M5S 3H8 (Canada); Ott, Christian D.; Scheel, Mark A.; Szilagyi, Bela [TAPIR, MC 350-17, California Institute of Technology, Pasadena, CA 91125 (United States); Kidder, Lawrence E.; Muhlberger, Curran D., E-mail: mbdeaton@wsu.edu, E-mail: m.duez@wsu.edu [Center for Radiophysics and Space Research, Cornell University, Ithaca, NY 14853 (United States)

    2013-10-10

    Neutrino emission significantly affects the evolution of the accretion tori formed in black hole-neutron star mergers. It removes energy from the disk, alters its composition, and provides a potential power source for a gamma-ray burst. To study these effects, simulations in general relativity with a hot microphysical equation of state (EOS) and neutrino feedback are needed. We present the first such simulation, using a neutrino leakage scheme for cooling to capture the most essential effects and considering a moderate mass (1.4 M{sub ☉} neutron star, 5.6 M{sub ☉} black hole), high-spin (black hole J/M {sup 2} = 0.9) system with the K{sub 0} = 220 MeV Lattimer-Swesty EOS. We find that about 0.08 M{sub ☉} of nuclear matter is ejected from the system, while another 0.3 M{sub ☉} forms a hot, compact accretion disk. The primary effects of the escaping neutrinos are (1) to make the disk much denser and more compact, (2) to cause the average electron fraction Y{sub e} of the disk to rise to about 0.2 and then gradually decrease again, and (3) to gradually cool the disk. The disk is initially hot (T ∼ 6 MeV) and luminous in neutrinos (L{sub ν} ∼ 10{sup 54} erg s{sup –1}), but the neutrino luminosity decreases by an order of magnitude over 50 ms of post-merger evolution.

  15. The 'practical elimination' approach of accident situations for water-cooled nuclear power reactors 2017

    International Nuclear Information System (INIS)

    2017-01-01

    The implementation of the defence in depth principle and current regulations have lead applicants to define provisions to prevent accidents, including severe accidents, and to limit their consequences should they occur. However, while defining the design orientations for a new water-cooled power reactor, applicants shall use the 'practical elimination' approach for severe accident situations (in the reactor core or the spent fuel pool) potentially leading to large early radiological releases, where it appears impossible to define realistic and demonstrable provisions to limit their consequences according to current knowledge and the techniques available at the time. The use of this approach should be discussed between the applicant and the safety authority at the design orientations stage; the authority will specify on a case-by-case basis the conditions for its approval. In order to 'practically eliminate' a situation, the designer shall first examine the possibility for making it physically impossible. Where physical impossibility cannot be achieved, provisions shall be implemented to justify with a high degree of confidence that the situation is extremely unlikely. Situations likely to be 'practically eliminated' are diverse (massive and rapid reactivity insertion accidents, explosions, containment bypasses, etc.); the justification of 'practical elimination' can only be assessed on a case-by-case basis, using deterministic considerations complemented by a probabilistic analysis. The assessment relies on the reactor physical characteristics as well as on the robustness and reliability of the lines of defence implemented to prevent the situation to be 'practically eliminated'. The implemented provisions shall be subject to strong design, manufacturing and operation requirements; considerations related to human factors and hazards shall also be taken into account. This document is an orientation text which defines

  16. Field testing of behavioral barriers for cooling water intake structures -test site 1 - Pickering Nuclear Generating Station

    International Nuclear Information System (INIS)

    Patrick, P.H.; McKinley, R.S.; Micheletti, W.C.

    1988-01-01

    A multi-year research program was developed by the Electric Power Research Institute to evaluate the effectiveness of selected behavioral systems for fish exclusion at sites representative of different aquatic environments. The first test site was the Pickering Nuclear Generating Station (NGS) located on Lake Ontario which represented the Great Lakes environment. A single pneumatic popper, a low frequency, high amplitude sound deterrent, was found to effectively exclude adult alewife, the principal species impinged at Pickering NGS. An air bubble curtain, used either alone or combined with strobe lights, was not a consistent deterrent. Effectiveness of air bubbles was only enhanced when used in association with a popper. Strobe lights were the least effective of the three devices tested. Operation of all three devices together did not surpass the effectiveness of the popper when used alone. Sound deterrents show promise for fish exclusion at generating stations located on the Great Lakes

  17. Analyses and computer code developments for accident-induced thermohydraulic transients in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Wulff, W.

    1977-01-01

    A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment

  18. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    International Nuclear Information System (INIS)

    1980-01-01

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050 0 C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850 0 C on several experimental alloys are discussed

  19. A Critical Review of the Recent Improvements in Minimizing Nuclear Waste by Innovative Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    E. Bomboni

    2008-01-01

    Full Text Available This paper presents a critical review of the recent improvements in minimizing nuclear waste in terms of quantities, long-term activities, and radiotoxicities by innovative GCRs, with particular emphasis to the results obtained at the University of Pisa. Regarding these last items, in the frame of some EU projects (GCFR, PUMA, and RAPHAEL, we analyzed symbiotic fuel cycles coupling current LWRs with HTRs, finally closing the cycle by GCFRs. Particularly, we analyzed fertile-free and Pu-Th-based fuel in HTR: we improved plutonium exploitation also by optimizing Pu/Th ratios in the fuel loaded in an HTR. Then, we chose GCFRs to burn residual MA. We have started the calculations on simplified models, but we ended them using more “realistic” models of the reactors. In addition, we have added the GCFR multiple recycling option using keff calculations for all the reactors. As a conclusion, we can state that, coupling HTR with GCFR, the geological disposal issues concerning high-level radiotoxicity of MA can be considerably reduced.

  20. ELECTRON COOLING FOR RHIC

    International Nuclear Information System (INIS)

    BEN-ZVI, I.; AHRENS, L.; BRENNAN, M.; HARRISON, M.; KEWISCH, J.; MACKAY, W.; PEGGS, S.; ROSER, T.; SATOGATA, T.; TRBOJEVIC, D.; YAKIMENKO, V.

    2001-01-01

    We introduce plans for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This project has a number of new features as electron coolers go: It will cool 100 GeV/nucleon ions with 50 MeV electrons; it will be the first attempt to cool a collider at storage-energy; and it will be the first cooler to use a bunched beam and a linear accelerator as the electron source. The linac will be superconducting with energy recovery. The electron source will be based on a photocathode gun. The project is carried out by the Collider-Accelerator Department at BNL in collaboration with the Budker Institute of Nuclear Physics

  1. Release of radioactive materials from nuclear power plants. Report No. 2. Dispersion mechanisms, transport paths, and concentration factors for radionuclides in the cooling water recipient

    International Nuclear Information System (INIS)

    Rygg, B.

    The discharge of radioactive materials in the cooling water from a nuclear power plant may involve consequences for the interests involved in the recipient and its organisms. Of special interest is the transport of radionuclides in water, sediments, and organisms to man. The most important elements are H, Na, P, Cr, Mn, Fe, Co, Zn, Sr, Mo, Ru, I, Cs, and Ce. Metals with high affinity for organic material will be sorbed to sediments rich in organic material, while other elements will be enriched in algaes and arrive in the sediment through decay or excrement. Elements in particulate form will normally precipitate. Ions will generally not be enriched in sediments. Marine organisms may take up nuclides directly from the water and from food. The concentration factor is dependent on the chemical properties of the element and the physiology of the organism. The occurrence of elements in water and organisms in the Oslofjord district is poorly known and tables have therefore been derived from literature data to indicate the concentration factors to be expected

  2. First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

    Science.gov (United States)

    You, Yan; Yoshida, Katsumi; Yano, Toyohiko

    2018-05-01

    Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

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

    Science.gov (United States)

    Guidez, Joel; Saturnin, Anne

    2017-11-01

    During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  4. Study of the influence of temperature and time on the electroplating nickel layer in Inconel 718 strips used in spacer grid of Pressurized Water Cooled nuclear reactors (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Renato; Abati, Amanda; Verne, Júlio; Panossian, Zehbour, E-mail: amanda.abati@marinha.mil.br, E-mail: jvernegropp@gmail.com, E-mail: renato.rezende@marinha.mil.br, E-mail: zep@ipt.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil). Laboratório de Desenvolvimento e Instrumentação de Combustível Nuclear; Instituto de Pesquisas Tecnológicas (IPT), São Paulo, SP (Brazil)

    2017-07-01

    The Inconel 718 (UNS N07718: Ni-{sup 19}Cr-{sup 18}Fe-{sup 5}Nb-3 Mo) is a precipitation hardenable nickel alloy that has good corrosion resistance and high mechanical strength. These strips are used for assembling the spacer grid of fuel element of pressurized water cooled nuclear reactors (PWR). The spacer grid is a structural component of fundamental importance in fuel elements of PWR reactors, maintaining the position and necessary spacing of the fuel rods within the arrangement of the fuel element. The spacer grid is formed by joining the points of intersection of the strips, by a joint process called brazing. For this process, these strips are stamped and plated with a thin layer of nickel by means of electroplating in order to protect against oxidation and allow a better flowability and wettability of the addition metal in the strips during brazing. Oxidation at the surface of the base material harms wettability and inhibits spreading of the liquid addition metal on the substrate surface during the brazing process. The use of coatings such as nickel plating is used to ensure such conditions. The results showed that there is a process of diffusion de some chemical elements such as chromium, iron, titanium and aluminum from the substrate to the nickel layer and nickel from the layer to the substrate. These chemical elements are responsible for the oxidation at the surface of the strip. (author)

  5. Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

    International Nuclear Information System (INIS)

    Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

    2005-01-01

    The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

  6. Homogenization and two scales convergence of some stationary and non-stationary heat transfer problems, application to gas cooled nuclear rectors

    International Nuclear Information System (INIS)

    Habibi, Z.

    2011-01-01

    We are interested in the homogenization of heat transfer in periodic porous media modelling the geometry of a gas cooled nuclear reactor. This geometry is made of a solid media perforated by several long thin parallel cylinders, the diameter of which is of the same order than the period. The heat is transported by conduction in the solid part of the domain and by conduction, convection and radiative transfer in the fluid part (the cylinders). A non-local boundary condition models the radiative heat transfer on the cylinder walls. It is a stationary analysis corresponding to a nominal performance of the reactor core, and also non-stationary corresponding to a normal shut-down of the core. To obtain the homogenized problem we first use a formal two-scale asymptotic expansion method. The mathematical justification of our results is based on the notion of two-scale convergence. One feature of this work in dimension 3 is that it combines homogenization with a 3D to 2D asymptotic analysis since the radiative transfer in the limit cell problem is purely two-dimensional. A second feature of this work is the study of this heat transfer when it contains an oscillating thermal source at the microscopic level and a thermal exchange with the perforations. In this context, our numerical analysis shows a non-negligible contribution of the second order corrector which helps us to model the gradients appearing between the source area and the perforations. (author) [fr

  7. Distribution of a pelagic tunicate, Salpa fusiformis in warm surface current of the eastern Korean waters and its impingement on cooling water intakes of Uljin nuclear power plant.

    Science.gov (United States)

    Chae, Jinho; Choi, Hyun Woo; Lee, Woo Jin; Kim, Dongsung; Lee, Jae Hac

    2008-07-01

    Impingement of a large amount of gelatinous plankton, Salpa fusiformis on the seawater intake system-screens in a nuclear power plant at Uljin was firstly recorded on 18th June 2003. Whole amount of the clogged animals was estimated were presumptively at 295 tons and the shortage of cooling seawater supply by the animal clogging caused 38% of decrease in generation capability of the power plant. Zooplankton collection with a multiple towing net during the day and at night from 5 to 6 June 2003 included various gelatinous zooplanktons known to be warm water species such as salps and siphonophores. Comparatively larger species, Salpa fusiformis occupied 25.4% in individual density among the gelatinous plankton and showed surface distribution in the depth shallower than thermocline, performing little diel vertical migration. Temperature, salinity and satellite data also showed warm surface current predominated over the southern coastal region near the power plant in June. The results suggested that warm surface current occasionally extended into the neritic region may transfer S. fusiformis, to the waters off the power plant. The environmental factors and their relation to ecobiology of the large quantity of salpa population that are being sucked into the intake channel of the power plant are discussed.

  8. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state......-of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  9. 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

  10. Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

    Energy Technology Data Exchange (ETDEWEB)

    Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

    2015-06-15

    In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the

  11. Forced convection heat transfer in rectangular ducts - general case of wall resistances and peripheral conduction for ventilation cooling of nuclear waste repositories [ heat transfer and nuclear disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lyczkowski, R. W.; Solbrig, C. W.; Gidaspow, D.

    1980-01-01

    A numerical solution for laminar flow heat transfer between a flowing gas and its containing rectangular duct has been obtained for many different boundary conditions which may arise in nuclear waste repository ventilation corridors. The problem has been solved for the cases of insulation on no walls, one wall, two walls, and three walls with various finite resistances on the remaining walls. Simplifications are made to decouple the convective heat transfer problem for the far field conduction problem, but peripheral conduction is retained. Results have been obtained for several duct aspect ratios in the thermal entrance and in the fully developed regions, including the constant temperature cases. When one wall is insulated and the other three are at constant temperature, the maximum temperature occurs in the fluid rather than on the insulated wall. This maximum moves toward the insulated wall with increasing axial distance. Nusselt numbers for the same constant flux on all four walls with peripheral conduction lie in a narrow band bounded by zero and infinite peripheral conduction cases. A dimensionless wall conduction group of four can be considered infinite for the purpose of estimating fully developed Nusselt numbers to within an accuracy of 3%. A decrease in wall and bulk temperatures by finite wall conduction has been demonstrated for the case of a black body radiation boundary condition. Nusselt numbers for the case of constant temperature on the top and bottom walls and constant heat flux on the side walls exhibited unexpected behavior.

  12. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  13. Cooling towers

    International Nuclear Information System (INIS)

    Korik, L.; Burger, R.

    1992-01-01

    What is the effect of 0.6C (1F) temperature rise across turbines, compressors, or evaporators? Enthalpy charts indicate for every 0.6C (1F) hotter water off the cooling tower will require an additional 2 1/2% more energy cost. Therefore, running 2.2C (4F) warmer due to substandard cooling towers could result in a 10% penalty for overcoming high heads and temperatures. If it costs $1,250,000.00 a year to operate the system, $125,000.00 is the energy penalty for hotter water. This paper investigates extra fuel costs involved in maintaining design electric production with cooling water 0.6C (1F) to 3C (5.5F) hotter than design. If design KWH cannot be maintained, paper will calculate dollar loss of saleable electricity. The presentation will conclude with examining the main causes of deficient cold water production. State-of-the-art upgrading and methodology available to retrofit existing cooling towers to optimize lower cooling water temperatures will be discussed

  14. Cooling tower

    Energy Technology Data Exchange (ETDEWEB)

    Norbaeck, P; Heneby, H

    1976-01-22

    Cooling towers to be transported on road vehicles as a unit are not allowed to exceed certain dimensions. In order to improve the efficiency of such a cooling tower (of cross-flow design and box-type body) with given dimensions, it is proposed to arrange at least one of the scrubbing bodies displaceable within a module or box. Then it can be moved out of the casing into working position, thereby increasing the front surface available for the inlet of air (and with it the efficiency) by nearly a factor of two.

  15. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; Lingerfelt, Eric J.; Wojtowicz, Anna

    2015-01-01

    Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

  16. Some problems raised by the operation of large nuclear turbo-generator sets. Cooling of shielded conductors. Precautionary steps for their thermal dimensioning

    International Nuclear Information System (INIS)

    Nisol, R.

    1976-01-01

    The role of the shielded conductors as power dissipation feeder from the generator towards the network is recalled. Their natural cooling limits and the possibilities of forced cooling are examined. The known incidence of short-circuit currents upon the components of the generator-network connection is reported [fr

  17. Modelization of cooling system components

    Energy Technology Data Exchange (ETDEWEB)

    Copete, Monica; Ortega, Silvia; Vaquero, Jose Carlos; Cervantes, Eva [Westinghouse Electric (Spain)

    2010-07-01

    In the site evaluation study for licensing a new nuclear power facility, the criteria involved could be grouped in health and safety, environment, socio-economics, engineering and cost-related. These encompass different aspects such as geology, seismology, cooling system requirements, weather conditions, flooding, population, and so on. The selection of the cooling system is function of different parameters as the gross electrical output, energy consumption, available area for cooling system components, environmental conditions, water consumption, and others. Moreover, in recent years, extreme environmental conditions have been experienced and stringent water availability limits have affected water use permits. Therefore, modifications or alternatives of current cooling system designs and operation are required as well as analyses of the different possibilities of cooling systems to optimize energy production taking into account water consumption among other important variables. There are two basic cooling system configurations: - Once-through or Open-cycle; - Recirculating or Closed-cycle. In a once-through cooling system (or open-cycle), water from an external water sources passes through the steam cycle condenser and is then returned to the source at a higher temperature with some level of contaminants. To minimize the thermal impact to the water source, a cooling tower may be added in a once-through system to allow air cooling of the water (with associated losses on site due to evaporation) prior to returning the water to its source. This system has a high thermal efficiency, and its operating and capital costs are very low. So, from an economical point of view, the open-cycle is preferred to closed-cycle system, especially if there are no water limitations or environmental restrictions. In a recirculating system (or closed-cycle), cooling water exits the condenser, goes through a fixed heat sink, and is then returned to the condenser. This configuration

  18. Stochastic cooling

    International Nuclear Information System (INIS)

    Bisognano, J.; Leemann, C.

    1982-03-01

    Stochastic cooling is the damping of betatron oscillations and momentum spread of a particle beam by a feedback system. In its simplest form, a pickup electrode detects the transverse positions or momenta of particles in a storage ring, and the signal produced is amplified and applied downstream to a kicker. The time delay of the cable and electronics is designed to match the transit time of particles along the arc of the storage ring between the pickup and kicker so that an individual particle receives the amplified version of the signal it produced at the pick-up. If there were only a single particle in the ring, it is obvious that betatron oscillations and momentum offset could be damped. However, in addition to its own signal, a particle receives signals from other beam particles. In the limit of an infinite number of particles, no damping could be achieved; we have Liouville's theorem with constant density of the phase space fluid. For a finite, albeit large number of particles, there remains a residue of the single particle damping which is of practical use in accumulating low phase space density beams of particles such as antiprotons. It was the realization of this fact that led to the invention of stochastic cooling by S. van der Meer in 1968. Since its conception, stochastic cooling has been the subject of much theoretical and experimental work. The earliest experiments were performed at the ISR in 1974, with the subsequent ICE studies firmly establishing the stochastic cooling technique. This work directly led to the design and construction of the Antiproton Accumulator at CERN and the beginnings of p anti p colliding beam physics at the SPS. Experiments in stochastic cooling have been performed at Fermilab in collaboration with LBL, and a design is currently under development for a anti p accumulator for the Tevatron

  19. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  20. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water; Estudio de un ensamble de combustible para el reactor nuclear de generacion IV enfriado con agua supercritica

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Espinosa P, G., E-mail: albrm29@yahoo.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, Av. San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico D. F. (MX)

    2011-11-15

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  1. ELECTRON COOLING FOR RHIC

    International Nuclear Information System (INIS)

    Ben-Zvi, I.

    2001-01-01

    The Accelerator Collider Department (CAD) at Brookhaven National Laboratory is operating the Relativistic Heavy Ion Collider (RHIC), which includes the dual-ring, 3.834 km circumference superconducting collider and the venerable AGS as the last part of the RHIC injection chain. CAD is planning on a luminosity upgrade of the machine under the designation RHIC II. One important component of the RHIC II upgrade is electron cooling of RHIC gold ion beams. For this purpose, BNL and the Budker Institute of Nuclear Physics in Novosibirsk entered into a collaboration aimed initially at the development of the electron cooling conceptual design, resolution of technical issues, and finally extend the collaboration towards the construction and commissioning of the cooler. Many of the results presented in this paper are derived from the Electron Cooling for RHIC Design Report [1], produced by the, BINP team within the framework of this collaboration. BNL is also collaborating with Fermi National Laboratory, Thomas Jefferson National Accelerator Facility and the University of Indiana on various aspects of electron cooling

  2. ELECTRON COOLING FOR RHIC.

    Energy Technology Data Exchange (ETDEWEB)

    BEN-ZVI,I.

    2001-05-13

    The Accelerator Collider Department (CAD) at Brookhaven National Laboratory is operating the Relativistic Heavy Ion Collider (RHIC), which includes the dual-ring, 3.834 km circumference superconducting collider and the venerable AGS as the last part of the RHIC injection chain. CAD is planning on a luminosity upgrade of the machine under the designation RHIC II. One important component of the RHIC II upgrade is electron cooling of RHIC gold ion beams. For this purpose, BNL and the Budker Institute of Nuclear Physics in Novosibirsk entered into a collaboration aimed initially at the development of the electron cooling conceptual design, resolution of technical issues, and finally extend the collaboration towards the construction and commissioning of the cooler. Many of the results presented in this paper are derived from the Electron Cooling for RHIC Design Report [1], produced by the, BINP team within the framework of this collaboration. BNL is also collaborating with Fermi National Laboratory, Thomas Jefferson National Accelerator Facility and the University of Indiana on various aspects of electron cooling.

  3. Passive cooling containment study

    International Nuclear Information System (INIS)

    Shin, J.J.; Iotti, R.C.; Wright, R.F.

    1993-01-01

    Pressure and temperature transients of nuclear reactor containment following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally for the full scale NPR (New Production Reactor). All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified CONTEMPT code CONTEMPT-NPR. In order to verify proper heat transfer coefficient, temperature, heat flux, and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and half foot (1.7m) diameter outer cylinder. Comparison of CONTEMPT-NPR and WGOTHIC was done for reduced scale NPR

  4. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  5. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  6. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  7. Passive containment cooling water distribution device

    Science.gov (United States)

    Conway, Lawrence E.; Fanto, Susan V.

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

  8. Cooling pancakes

    International Nuclear Information System (INIS)

    Bond, J.R.; Wilson, J.R.

    1984-01-01

    In theories of galaxy formation with a damping cut-off in the density fluctuation spectrum, the first non-linear structures to form are Zeldovich pancakes in which dissipation separates gas from any collisionless dark matter then present. One-dimensional numerical simulations of the collapse, shock heating, and subsequent thermal evolution of pancakes are described. Neutrinos (or any other cool collisionless particles) are followed by direct N-body methods and the gas by Eulerian hydrodynamics with conduction as well as cooling included. It is found that the pressure is relatively uniform within the shocked region and approximately equals the instantaneous ram pressure acting at the shock front. An analytic theory based upon this result accurately describes the numerical calculations. (author)

  9. Cool Sportswear

    Science.gov (United States)

    1982-01-01

    New athletic wear design based on the circulating liquid cooling system used in the astronaut's space suits, allows athletes to perform more strenuous activity without becoming overheated. Techni-Clothes gear incorporates packets containing a heat-absorbing gel that slips into an insulated pocket of the athletic garment and is positioned near parts of the body where heat transfer is most efficient. A gel packet is good for about one hour. Easily replaced from a supply of spares in an insulated container worn on the belt. The products, targeted primarily for runners and joggers and any other athlete whose performance may be affected by hot weather, include cooling headbands, wrist bands and running shorts with gel-pack pockets.

  10. Cooling systems

    International Nuclear Information System (INIS)

    Coutant, C.C.

    1978-01-01

    Progress on the thermal effects project is reported with regard to physiology and distribution of Corbicula; power plant effects studies on burrowing mayfly populations; comparative thermal responses of largemouth bass from northern and southern populations; temperature selection by striped bass in Cherokee Reservoir; fish population studies; and predictive thermoregulation by fishes. Progress is also reported on the following; cause and ecological ramifications of threadfin shad impingement; entrainment project; aquaculture project; pathogenic amoeba project; and cooling tower drift project

  11. The fixed bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Sahin, S.; Sefidvash, F.

    2007-01-01

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

  12. Fundamental approaches to predicting stress corrosion: 'Quantitative micro-nano' (QMN) approach to predicting stress corrosion cracking in water cooled nuclear plants

    International Nuclear Information System (INIS)

    Staehle, R.W.

    2010-01-01

    This paper describes the modeling and experimental studies of stress corrosion cracking with full disciplinary set at the atomic level. Its objective is to develop an intellectual structure for quantitative prediction of stress corrosion cracking in water cooled reactors.

  13. The influence of atmospheric conditions on the cooling tower plume of nuclear power station; Uticaj atmosferskih uslova na perjanicu rashladnih tornjeva nuklearne elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Vehauc, A [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    The paper deals with the effect of atmospheric conditions - relative humidity, wind velocity, temperature and temperature gradient on the visible plume. For estimating cooling tower plumes, used was made of verified mathematical model. (author)

  14. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Roguin, A. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de

  15. Can-rupture detection in gas-cooled nuclear reactors; La detection des ruptures de gaine dans les piles nucleaires refroidies par gaz

    Energy Technology Data Exchange (ETDEWEB)

    Roguin, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    Can-rupture detection (DRG) is one important aspect of pile safety, more particularly so in the case of gas-cooled reactors. A rapid and sure detection constitutes also an improvement as far as the efficiency of electricity-producing nuclear power stations are concerned. Among the numerous can-rupture detection methods, that based on the measurement of the concentration of short-lived fission gases in the heat-carrying fluid has proved to be the most sensitive and the most rapid. A systematic study of detectors based on the electrostatic collection of the daughter products of fission gases has been undertaken with a view to equip the reactors EL 2, G 3, EDF 1, EDF 2 and EDF 3, the gas loops of PEGASE and EL 4. The different parameters are studied in detail in order to obtain a maximum sensitivity and to make it possible to construct detection devices having the maximum operational reliability and requiring the minimum maintenance. The primary applications of these devices are examined in the case of the above-mentioned reactors. (author) [French] La Detection des Ruptures de Gaines (D. R. G.) est un aspect important de la securite des piles et plus particulierement des piles refroidies par un gaz. Une detection rapide et sure constitue aussi un element d'amelioration du rendement des centrales nucleaires productrices d'energie electrique. Parmi les nombreuses methodes de detection des ruptures de gaines, la mesure de la concentration dans le fluide caloporteur des gaz de fission a vie courte s'est revelee comme la plus sensible et la plus rapide. Une etude systematique des detecteurs a collection electrostatique des descendants des gaz de fission a ete entreprise en vue d'equiper les piles EL 2, G 3, EDF 1, EDF 2 et EDF 3, les boucles a gaz de la pile Pegase et la pile EL 4. Les divers parametres sont etudies en detail pour obtenir une sensibilite maximum et permettre la realisation de dispositifs de detection ayant le maximum de securite de fonctionnement et le

  16. Development of Probabilistic Safety Assessment with respect to the first demonstration nuclear power plant of high temperature gas cooled reactor in China

    International Nuclear Information System (INIS)

    Tong Jiejuan; Zhao Jun; Liu Tao; Xue Dazhi

    2012-01-01

    Due to the unique concept of HTR-PM (High Temperature Gas Cooled Reactor-Pebble Bed Module) design, Chinese nuclear authority has anticipated that HTR-PM will bring challenge to the present regulation. The pilot use of PSA (Probabilistic Safety Assessment) during HTR-PM design and safety review is deemed to be the necessary and efficient tool to tackle the problem, and is actively encouraged as indicated in the authority's specific policy statement on HTR-PM project. The paper summarizes the policy statement to set up the base of PSA development and application activities. The up-to-date status of HTR-PM PSA development and the risk-informed application activities are introduced in this paper as the follow-up response to the policy statement. For open discussion, the paper hereafter puts forward several technical issues which have been encountered during HTR-PM PSA development. Since HTR-PM PSA development experience has the general conclusion that many of the PSA elements can be and have been implemented successfully by the traditional PSA techniques, only the issues which extra innovative efforts may be needed are highlighted in this paper. They are safety goal and risk metrics, PSA modeling framework for the non-water reactors, passive system reliability evaluation, initiating events frequencies and component reliability data estimation techniques for the new reactors and so on. The paper presents the way in which the encountered technical issues were or will be solved, although the proposed way may not be the ultimate best solution. The paper intends to express the standpoint that although the PSA of new reactor has the inherent weakness due to the insufficient information and larger data uncertainty, the problem of component reliability data is much less severe than people have conceived. The unique design conception and functional features of the reactors can influence the results more significantly than the component reliability data. What we are benefited

  17. Updating of ASME Nuclear Code Case N-201 to Accommodate the Needs of Metallic Core Support Structures for High Temperature Gas Cooled Reactors Currently in Development

    International Nuclear Information System (INIS)

    Basol, Mit; Kielb, John F.; MuHooly, John F.; Smit, Kobus

    2007-01-01

    On September 29, 2005, ASME Standards Technology, LLC (ASME ST-LLC) executed a multi-year, cooperative agreement with the United States DOE for the Generation IV Reactor Materials project. The project's objective is to update and expand appropriate materials, construction, and design codes for application in future Generation IV nuclear reactor systems that operate at elevated temperatures. Task 4 was embarked upon in recognition of the large quantity of ongoing reactor designs utilizing high temperature technology. Since Code Case N-201 had not seen a significant revision (except for a minor revision in September, 2006 to change the SA-336 forging reference for 304SS and 316SS to SA-965 in Tables 1.2(a) and 1.2(b), and some minor editorial changes) since December 1994, identifying recommended updates to support the current high temperature Core Support Structure (CSS) designs and potential new designs was important. As anticipated, the Task 4 effort identified a number of Code Case N-201 issues. Items requiring further consideration range from addressing apparent inconsistencies in definitions and certain material properties between CC-N-201 and Subsection NH, to inclusion of additional materials to provide the designer more flexibility of design. Task 4 developed a design parameter survey that requested input from the CSS designers of ongoing high temperature gas cooled reactor metallic core support designs. The responses to the survey provided Task 4 valuable input to identify the design operating parameters and future needs of the CSS designers. Types of materials, metal temperature, time of exposure, design pressure, design life, and fluence levels were included in the Task 4 survey responses. The results of the survey are included in this report. This research proves that additional work must be done to update Code Case N-201. Task 4 activities provide the framework for the Code Case N-201 update and future work to provide input on materials. Candidate

  18. Passive vibro-acoustic detection of a sodium-water reaction in a steam generator of a sodium-cooled fast neutrons nuclear reactor by beam forming

    International Nuclear Information System (INIS)

    Moriot, Jeremy

    2013-01-01

    This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow ≤ 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beam forming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beam forming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beam forming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydro-phone. The performances of the beam forming are studied for different water flow speeds and different amplitude and frequencies of the source. (author) [fr

  19. Replacement of the cooling tower packing at the Goesgen-Daeniken AG nuclear power plant; Ersatz der Kuehlturmeinbauten im Kernkraftwerk Goesgen-Daeniken

    Energy Technology Data Exchange (ETDEWEB)

    Rich, Hans Walter [Kernkraftwerk Goesgen-Daeniken AG, Daeniken (Switzerland)

    2012-07-01

    In 2005 the asbestos cement cooling tower packing was replaced by plastic material. Two years later, the packing showed strong deformations, deposits of solids and weight gain. At the end of 2007 parts of the packing collapsed into the cooling tower basin. Investigations were made, revealing that the thickness of the packing foil was too low and that packing geometry and biofilms on the surface of the packing favoured deposition of solids. Successful measures were taken to solve the problems. (orig.)

  20. Beam Cooling with ionisation losses

    CERN Document Server

    Rubbia, Carlo; Kadi, Y; Vlachoudis, V

    2006-01-01

    A novel type of particle "cooling", called Ionization Cooling, is applicable to slow (v of the order of 0.1c) ions stored in a small ring. The many traversals through a thin foil enhance the nuclear reaction probability, in a steady configuration in which ionisation losses are recovered at each turn by a RF-cavity. For a uniform target "foil" the longitudinal momentum spread diverges exponentially since faster (slower) particles ionise less (more) than the average. In order to "cool" also longitudinally, a chromaticity has to be introduced with a wedge shaped "foil". Multiple scattering and straggling are then "cooled" in all three dimensions, with a method similar to the one of synchrotron cooling, but valid for low energy ions. Particles then stably circulate in the beam indefinitely, until they undergo for instance nuclear processes in the thin target foil. This new method is under consideration for the nuclear production of a few MeV/A ion beams. Simple reactions, for instance Li 7 + D Li 8 + p, are more ...

  1. Nuclear decommissioning

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The paper on nuclear decommissioning was presented by Dr H. Lawton to a meeting of the British Nuclear Energy Society and Institution of Nuclear Engineers, 1986. The decommissioning work currently being undertaken on the Windscale advanced gas cooled reactor (WAGR) is briefly described, along with projects in other countries, development work associated with the WAGR operation and costs. (U.K.)

  2. Nuclear Reactor RA Safety Report, Vol. 5, Reactor cooling systems; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 5, Hladjenje reaktora i pripadajuci sistemi

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    RA reactor cooling system enable cooling during normal operation and under possible accidental conditions and include: technical water system, heavy water system, helium gas system, system for heavy water purification and emergency cooling system. Primary cooling system is a closed heavy water circulation system. Heavy water system is designed to enable permanent circulation and twofold function of heavy water. In the upward direction of cooling it has a coolant role and in the downward direction it is the moderator. Separate part of the primary coolant loop is the system for heavy water purification. This system uses distillation and ion exchange processes. [Serbo-Croat] Sistemi za hladjenje reaktora RA obezbedjuju hladjenje u svim rezimima eksploatacije i potenijalno udesnim satnjim i obuhvataju: sistem tehnicke vode, sistem teske vode, gasni sistem helijuma, sistem za preciscavanje teske vode i sistem za akcidentalno hladjenje. Primarni sistem za hladjenje je zatvoreni cirkulacioni sistem teske vode. Specificnost sistema teske vode vezana je za neprekidnost cirkulacije i razlicite funkcije teske vode. U uzlaznom strujanju, teske vode ima ulogu hladioca a u silaznom ulogu moderatora. Poseban deo primarnog sistema predstavlja sistem za preciscavanje teske vode. Ovaj sistem koristi destilacioni i jonoizmenjivacki postupak.

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

    International Nuclear Information System (INIS)

    2013-05-01

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

  4. Cooling of Accretion-Heated Neutron Stars

    Indian Academy of Sciences (India)

    Rudy Wijnands

    2017-09-12

    Sep 12, 2017 ... the magnetic field might play an important role in the heating and cooling of the neutron stars. .... Source near Sgr A ..... marked the start of the research field that uses the cool- ... This curve is just to guide the eye for the individual sources and it is clear ..... Not all accretion-induced nuclear reactions might.

  5. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

    Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

  6. Cool snacks

    DEFF Research Database (Denmark)

    Grunert, Klaus G; Brock, Steen; Brunsø, Karen

    2016-01-01

    Young people snack and their snacking habits are not always healthy. We address the questions whether it is possible to develop a new snack product that adolescents will find attractive, even though it is based on ingredients as healthy as fruits and vegetables, and we argue that developing...... such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...... people's snacking behaviour and then to develop and test new, healthier snacking solutions. These new snacking solutions were tested and found to be favourably accepted by young people. The paper therefore provides a proof of principle that the development of snacks that are both healthy and attractive...

  7. Cool visitors

    CERN Multimedia

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  8. Penning-trap mass spectrometry of radioactive, highly charged ions. Measurements of neutron-rich Rb and Sr nuclides for nuclear astrophysics and development of a novel Penning trap for cooling highly charged ions

    International Nuclear Information System (INIS)

    Simon, Vanessa Veronique

    2012-01-01

    High-precision atomic mass measurements are vital for the description of nuclear structure, investigations of nuclear astrophysical processes, and tests of fundamental symmetries. The neutron-rich A ∼ 100 region presents challenges for modeling the astrophysical r-process because of sudden nuclear shape transitions. This thesis reports on high-precision masses of short-lived neutron-rich 94,97,98 Rb and 94,97-99 Sr isotopes using the TITAN Penning-trap mass spectrometer at TRIUMF. The isotopes were charge-bred to q = 15+; uncertainties of less than 4 keV were achieved. Results deviate by up to 11σ compared to earlier measurements and extend the region of nuclear deformation observed in the A∼100 region. A parameterized r-process model network calculation shows that mass uncertainties for the elemental abundances in this region are now negligible. Although beneficial for the measurement precision, the charge breeding process leads to an increased energy spread of the ions on the order of tens of eV/q. To eliminate this drawback, a Cooler Penning Trap (CPET) has been developed as part of this thesis. The novel multi-electrode trap structure of CPET forms nested potentials to cool HCI sympathetically using either electrons or protons to increase the overall efficiency and precision of the mass measurement. The status of the off-line setup and initial commissioning experiments are presented.

  9. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

    Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

  10. WORKSHOP: Beam cooling

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Cooling - the control of unruly particles to provide well-behaved beams - has become a major new tool in accelerator physics. The main approaches of electron cooling pioneered by Gersh Budker at Novosibirsk and stochastic cooling by Simon van der Meer at CERN, are now complemented by additional ideas, such as laser cooling of ions and ionization cooling of muons

  11. Nuclear

    International Nuclear Information System (INIS)

    2014-01-01

    This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

  12. 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

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

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

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

  14. Preliminary matrix model for quantifying and balancing the socio-economic impact of alternative cooling system technologies for nuclear power plants

    International Nuclear Information System (INIS)

    Cleary, D.P.; Salomon, S.N.; Pollnow, L.A.; Spangler, M.B.

    1976-01-01

    Assessment of environmental, including socio-economic, impacts of alternative technologies or courses of action is made difficult by the inability to adequately quantify the impacts. Matrix methods offer a set of techniques which allows the analyst to compare the relative impacts of alternative technologies or actions. Work is underway to develop and adapt these techniques to be used in assessing the environmental impacts of alternative cooling systems, and other alternative technological and siting options

  15. Automation of the electromagnetic filter applied for condensation water treatment in the secondary cooling system of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Szilagyi, Gyoergy

    1989-01-01

    A full-flow condensation water purification system is applied in the secondary cooling circuit of the Paks NPP. The electromagnetic filter of the filtering system eliminates ferromagnetic impurities. The filter consists of a high current coil and an automatic control unit. During the improvement of this unit, a FESTO FPC-404 type controller based on an extended capability PLC was installed. (R.P.) 5 figs

  16. Renewable Heating And Cooling

    Science.gov (United States)

    Renewable heating and cooling is a set of alternative resources and technologies that can be used in place of conventional heating and cooling technologies for common applications such as water heating, space heating, space cooling and process heat.

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  18. Thermodynamic analysis of the use a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water-splitting cycle for hydrogen production

    International Nuclear Information System (INIS)

    Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

    2008-01-01

    Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved 'steam' parameters (outlet temperatures up to 625degC and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600degC. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the 'nuclear' heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of

  19. Restaurant food cooling practices.

    Science.gov (United States)

    Brown, Laura Green; Ripley, Danny; Blade, Henry; Reimann, Dave; Everstine, Karen; Nicholas, Dave; Egan, Jessica; Koktavy, Nicole; Quilliam, Daniela N

    2012-12-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention's Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study.

  20. Emergency cooling apparatus for reactor

    International Nuclear Information System (INIS)

    Sakaguchi, S.

    1975-01-01

    A nuclear reactor is described which has the core surrounded by coolant and an inert cover gas all sealed within a container, an emergency cooling apparatus employing a detector that will detect cover gas or coolant, particularly liquid sodium, leaking from the container of the reactor, to release a heat exchange material that is inert to the coolant, which heat exchange material is cooled during operation of the reactor. The heat exchange material may be liquid niitrogen or a combination of spheres and liquid nitrogen, for example, and is introduced so as to contact the coolant that has leaked from the container quickly so as to rapidly cool the coolant to prevent or extinguish combustion. (Official Gazette)

  1. Emergency core cooling systems

    International Nuclear Information System (INIS)

    Kubokoya, Takashi; Okataku, Yasukuni.

    1984-01-01

    Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

  2. Assessment of spent fuel cooling

    International Nuclear Information System (INIS)

    Ibarra, J.G.; Jones, W.R.; Lanik, G.F.

    1997-01-01

    The paper presents the methodology, the findings, and the conclusions of a study that was done by the Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) on loss of spent fuel pool cooling. The study involved an examination of spent fuel pool designs, operating experience, operating practices, and procedures. AEOD's work was augmented in the area of statistics and probabilistic risk assessment by experts from the Idaho Nuclear Engineering Laboratory. Operating experience was integrated into a probabilistic risk assessment to gain insight on the risks from spent fuel pools

  3. Evolution of the thickness of the aluminum oxide film due to the pH of the cooling water and surface temperature of the fuel elements clad of a nuclear reactor

    International Nuclear Information System (INIS)

    Babiche, Ivan

    2013-01-01

    This paper describes the mechanism of growth of a film of aluminum oxide on an alloy of the same material, which serves as a protective surface being the constituent material of the RP-10 nuclear reactor fuel elements clads. The most influential parameters on the growth of this film are: the pH of the cooling water and the clad surface temperature of the fuel element. For this study, a mathematical model relating the evolution of the aluminum oxide layer thickness over the time, according to the same oxide film using a power law is used. It is concluded that the time of irradiation, the heat flux at the surface of the aluminum material, the speed of the coolant, the thermal conductivity of the oxide, the initial thickness of the oxide layer and the solubility of the protective oxide are parameters affecting in the rate and film formation. (author).

  4. Improvements in the design of the engineered safeguards cooling water systems at Vandellos II nuclear power plant.; Mejoras del diseno de los sistemas de agua de refrigeracion de salvaguardias tecnologicas de la central nuclear Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    Barrado, M.

    2009-07-01

    The investigation performed following the essential services water system incident that occurred in 2004 underlined the need to improve the design and management of the plants cooling water systems, and the decision was taken to create a new system, known as EJ, which has recently entered service. The characteristics of this system are explained in this article. (Author)

  5. Cooled Water Production System,

    Science.gov (United States)

    The invention refers to the field of air conditioning and regards an apparatus for obtaining cooled water . The purpose of the invention is to develop...such a system for obtaining cooled water which would permit the maximum use of the cooling effect of the water -cooling tower.

  6. Hybrid radiator cooling system

    Science.gov (United States)

    France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

    2016-03-15

    A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

  7. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  8. A method and device for cooling

    International Nuclear Information System (INIS)

    Gautier, Daniel.

    1974-01-01

    The invention relates to a method and a device for cooling steam. The invention refers to a method for cooling steam from a turbine, e.g. a turbine coupled to a high power nuclear reactor, in which a fluid F in the state of a two-phase mixture (steam and condensation liquid) is circulated, in a closed circuit, in conduits passing through a condenser associated with the turbine in which fluid F contained in said conduits vaporizes by heat-exchange with the condenser hot steam, then through a cooling tower wherein the fluid condenses by heat-exchange with a coolant. This can be applied to cooling the steam in a turbine associated with a nuclear reactor [fr

  9. Nuclear astrophysics

    International Nuclear Information System (INIS)

    Haxton, W.C.

    1992-01-01

    The problem of core-collapse supernovae is used to illustrate the many connections between nuclear astrophysics and the problems nuclear physicists study in terrestrial laboratories. Efforts to better understand the collapse and mantle ejection are also motivated by a variety of interdisciplinary issues in nuclear, particle, and astrophysics, including galactic chemical evolution, neutrino masses and mixing, and stellar cooling by the emission of new particles. The current status of theory and observations is summarized

  10. Dry cooling systems with plastic surfaces

    International Nuclear Information System (INIS)

    Roma, Carlo; Leonelli, Vincenzo

    1975-01-01

    Research and experiments made on dry cooling systems with plastic surfaces are described. The demonstration program planned in Italy for a 100Gcal/h dry cooling system is exposed, and an installation intended for a large 1300Mwe nuclear power station is described with reference to the assembly (exploitation and maintenance included). The performance and economic data relating to this installation are also exposed [fr

  11. Kinetic model for predicting the composition of chlorinated water discharged from power plant cooling systems

    International Nuclear Information System (INIS)

    Lietzke, M.H.

    1977-01-01

    The results of applying a kinetic model to the chlorination data supplied by Commonwealth Edison on the once-through cooling system at the Quad Cities Nuclear Station provide a validation of the model. The two examples given demonstrate that the model may be applied to either once-through cooling systems or to cooling systems involving cooling towers

  12. 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

  13. Restaurant Food Cooling Practices†

    Science.gov (United States)

    BROWN, LAURA GREEN; RIPLEY, DANNY; BLADE, HENRY; REIMANN, DAVE; EVERSTINE, KAREN; NICHOLAS, DAVE; EGAN, JESSICA; KOKTAVY, NICOLE; QUILLIAM, DANIELA N.

    2017-01-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention’s Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study. PMID:23212014

  14. Control of microstructure in brazed zone of Zircaloy-4 nuclear fuel sheathing by optimization of Σ(C+P+Si) contents and cooling schedules

    International Nuclear Information System (INIS)

    Quach, V.; Northwood, D.O.

    1985-01-01

    In the production of fuel elements for the CANDU-PHW reactor, induction brazing is used to attach appendages (bearing and split spacer pads) onto the outside wall of the Zircaloy-4 sheathing. The brazing process, 40 to 60 seconds at temperature in excess of 1000 0 C, produces 3 heat-affected zones amounting to about 30% of the thickness. These heat affected zones quite often contain large grains and either a basketweave or a parallel plate type of Widmanstatten structure. Small grains and a basketweave structure are preferred. Using simulated brazing treatments, it is demonstrated that by control of the impurity content, Σ(C+P+Si), and cooling rate from the brazing temperature, the desired microstructure can be obtained in the braze heat-affected zone. The formation of the basketweave structure is promoted by higher impurity contents, with the second phase impurity particles acting as nuclei for the basketweave structure in preference to the β-grain boundaries where the parallel plate structure is nucleated

  15. Radiative heat transfer in the Na mist dispersion over the hot surface of liquid Na in the cooling system of nuclear reactor

    International Nuclear Information System (INIS)

    Kunitomo, T.; Shafey, H.M.

    1980-01-01

    The analysis has been carried out for the radiative heat transfer in the Na mist dispersion enclosed between the hot surface of liquid Na at temperature Tsub(n) and the cold surface of Na deposit at Tsub(c). The model selected for the present study represents the Na mist formed in a sodium cooled fast breeder reactor in which the condensed liquid particles are dispersed in the mixture of the Ar cover gas and the Na vapor. The analysis is based on replacing the inhomogeneous dispersing medium by three discrete homogeneous layers, and formulating the transfer equation for the monochromatic radiation in each layer according to the Chandrasekhar theory. The numerical calculations of the radiative qsub(r) and convective qsub(c) heat transfers have been performed for the wave length range lambda=1.6-30 μm and are compared. The qsub(r) has the same order of magnitude as the qsub(c) for all conditions of the mist dispersions. Both qsub(r) and qsub(c) increase by nearly equal rates with the increase of Tsub(H) and decrease by different rates with increasing Tsub(c). Variations of the particle diameter of the Na mist do not change substantially the qsub(r). Both qsub(r) and qsub(c) decrease slightly with the increase in the total thickness of the Na mist dispersion

  16. Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

    2016-04-15

    Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

  17. Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

    International Nuclear Information System (INIS)

    Le-Guillou, Mael

    2014-01-01

    In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)

  18. Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

    International Nuclear Information System (INIS)

    Lemercier, G.; Lions, N.

    1986-01-01

    The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  20. Nuclear

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    The first text deals with a new circular concerning the collect of the medicine radioactive wastes, containing radium. This campaign wants to incite people to let go their radioactive wastes (needles, tubes) in order to suppress any danger. The second text presents a decree of the 31 december 1999, relative to the limitations of noise and external risks resulting from the nuclear facilities exploitation: noise, atmospheric pollution, water pollution, wastes management and fire prevention. (A.L.B.)