WorldWideScience

Sample records for nuclear heat transport

  1. Ultimate heat sink and directly associated heat transport systems for nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    The scope of the Guide covers design considerations for various types of ultimate heat sinks (UHS) and directly associated heat transport systems, and for types and sources of related heat transport fluids. The scope encompasses the conditions for using the UHS for reactor safety following postulated initiating events, as well as its selection, sizing and reliability

  2. DOS-HEATING6: A general conduction code with nuclear heat generation derived from DOT-IV transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M.L.; Yuecel, A.; Nadkarny, S.

    1988-05-01

    The HEATING6 heat conduction code is modified to (a) read the multigroup particle fluxes from a two-dimensional DOT-IV neutron- photon transport calculation, (b) interpolate the fluxes from the DOT-IV variable (optional) mesh to the HEATING6 control volume mesh, and (c) fold the interpolated fluxes with kerma factors to obtain a nuclear heating source for the heat conduction equation. The modified HEATING6 is placed as a module in the ORNL discrete ordinates system (DOS), and has been renamed DOS-HEATING6. DOS-HEATING6 provides the capability for determining temperature distributions due to nuclear heating in complex, multi-dimensional systems. All of the original capabilities of HEATING6 are retained for the nuclear heating calculation; e.g., generalized boundary conditions (convective, radiative, finned, fixed temperature or heat flux), temperature and space dependent thermal properties, steady-state or transient analysis, general geometry description, etc. The numerical techniques used in the code are reviewed and the user input instructions and JCL to perform DOS-HEATING6 calculations are presented. Finally a sample problem involving coupled DOT-IV and DOS-HEATING6 calculations of a complex space-reactor configurations described, and the input and output of the calculations are listed. 10 refs., 11 figs., 6 tabs.

  3. Advanced Intermediate Heat Transport Loop Design Configurations for Hydrogen Production Using High Temperature Nuclear Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Cliff Davis; Rober Barner; Paul Pickard

    2005-01-01

    The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic evaluations and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various

  4. Nuclear fuel transport flasks

    International Nuclear Information System (INIS)

    Burgess, M.H.; Fry, C.J.

    1984-01-01

    A nuclear fuel transport flask has a surrounding structure carrying inwardly directed heat transfer fins additional to the normal outwardly directed heat transfer fins on the main body of the flask. The additional fins can be interleaved with the main fins, and the structure carrying the additional fins can either be a shroud or an open framework. (author)

  5. Low temperature nuclear heat

    Energy Technology Data Exchange (ETDEWEB)

    Kotakorpi, J.; Tarjanne, R. (comps.)

    1977-08-01

    The meeting was concerned with the use of low grade nuclear heat for district heating, desalination, process heat, and agriculture and aquaculture. The sessions covered applications and demand, heat sources, and economics.

  6. Nuclear district heating

    International Nuclear Information System (INIS)

    Ricateau, P.

    1976-01-01

    An economic study of nuclear district heating is concerned with: heat production, its transmission towards the area to be served and the distribution management towards the consumers. Foreign and French assessments show that the high cost of now existing techniques of hot water transport defines the competing limit distance between the site and township to be below some fifty kilometers for the most important townships (provided that the fuel price remain stationary). All studies converge towards the choice of a high transport temperature as soon as the distance is of some twenty kilometers. As for fossile energy saving, some new possibilities appear with process heat reactors; either PWR of about 1000MWth for large townships, or pool-type reactors of about 100MWth when a combination with an industrial steam supply occurs [fr

  7. Heat resistant materials and their feasibility issues for a space nuclear transportation system

    International Nuclear Information System (INIS)

    Olsen, C.S.

    1991-01-01

    A number of nuclear propulsion concepts based on solid-core nuclear propulsion are being evaluated for a nuclear propulsion transportation system to support the Space Exploration Initiative (SEI) involving the reestablishment of a manned lunar base and the subsequent exploration of Mars. These systems will require high-temperature materials to meet the operating conditions with appropriate reliability and safety built into these systems through the selection and testing of appropriate materials. The application of materials for nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP) systems and the feasibility issues identified for their use will be discussed. Some mechanical property measurements have been obtained, and compatibility tests were conducted to help identify feasibility issues. 3 refs., 1 fig., 4 tabs

  8. Paleoclassical electron heat transport

    International Nuclear Information System (INIS)

    Callen, J.D.

    2005-01-01

    Radial electron heat transport in low collisionality, magnetically-confined toroidal plasmas is shown to result from paleoclassical Coulomb collision processes (parallel electron heat conduction and magnetic field diffusion). In such plasmas the electron temperature equilibrates along magnetic field lines a long length L, which is the minimum of the electron collision length and a maximum effective half length of helical field lines. Thus, the diffusing field lines induce a radial electron heat diffusivity M ≅ L/(πR 0q ) ∼ 10 >> 1 times the magnetic field diffusivity η/μ 0 ≅ ν e (c/ω p ) 2 . The paleoclassical electron heat flux model provides interpretations for many features of 'anomalous' electron heat transport: magnitude and radial profile of electron heat diffusivity (in tokamaks, STs, and RFPs), Alcator scaling in high density plasmas, transport barriers around low order rational surfaces and near a separatrix, and a natural heat pinch (or minimum temperature gradient) heat flux form. (author)

  9. Nuclear transport

    International Nuclear Information System (INIS)

    Anon.

    2001-01-01

    Here is given the decree (2001/1199) of the 10. of december 2001 relative to the passing of safety rules concerning the maritime transport of spent fuels, plutonium and high-level radioactive wastes contained in packages. (O.M.)

  10. Urban district heating using nuclear heat - a survey

    International Nuclear Information System (INIS)

    Beresovski, T.; Oliker, I.

    1979-01-01

    The use of heat from nuclear power plants is of great interest in connection with projected future expansions of large urban district heating systems. Oil price escalation and air pollution from increased burning of fossil fuels are substantial incentivers for the adoption of nuclear heat and power plants. The cost of the hot water piping system from the nuclear plant to the city is a major factor in determining the feasibility of using nuclear heat. To achieve reasonable costs, the heat load should be at least 1500 MW(th), transport temperatures 125-200 0 C and distances preferably 50 km or less. Heat may be extracted from the turbines of conventional power reactors. Alternatively, some special-purpose smaller reactors are under development which are specially suited to production of heat with little or no power coproduct. Many countries are conducting studies of future expansions of district heating systems to use nuclear heat. Several countries are developing technology suitable for this application. Actual experience with the use of nuclear heat for district heating is currently being gained only in the USSR, however. While district heating appears to be a desirable technology at a time of increasing fossil-fuel costs, the use of nuclear heat will require siting of nuclear plants within transmission radius of cities. The institutional barries toward use of nuclear heating will have to be overcome before the energy conservation potential of this approach can be realized on a significant scale. (author)

  11. Heat transport and storage

    International Nuclear Information System (INIS)

    Despois, J.

    1977-01-01

    Recalling the close connections existing between heat transport and storage, some general considerations on the problem of heat distribution and transport are presented 'in order to set out the problem' of storage in concrete form. This problem is considered in its overall plane, then studied under the angle of the different technical choices it involves. The two alternatives currently in consideration are described i.e.: storage in a mined cavity and underground storage as captive sheet [fr

  12. Optimum design of a nuclear heat supply

    International Nuclear Information System (INIS)

    Borel, J.P.

    1984-01-01

    This paper presents an economic analysis for the optimum design of a nuclear heat supply to a given district-heating network. First, a general description of the system is given, which includes a nuclear power plant, a heating power plant and a district-heating network. The heating power plant is fed with steam from the nuclear power plant. It is assumed that the heating network is already in operation and that the nuclear power plant was previously designed to supply electricity. Second, a technical definition of the heat production and transportation installations is given. The optimal power of these installations is examined. The main result is a relationship between the network capacity and the level of the nuclear heat supply as a substitute for oil under the best economic conditions. The analysis also presents information for choosing the best operating mode. Finally, the heating power plant is studied in more detail from the energy, technical and economic aspects. (author)

  13. Heat transport the cold way

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    A novel system for long-distance heat transport is being born in the 'Kernforschungsanlage Juelich' with the project being called 'Nukleare Fernenergie' (nuclear district energy). The project is also known as 'EVA/ADAM' [EVA = Einzelrohr-Versuchs-Anlage (single tube test facility); ADAM = Anlage mit Drei Adiabaten Methanisierungsreaktoren (plant provided with three adiabate methanising reactors)] and is based in principle on transport of energy in chemical bond within a closed loop. In the 60ies already this development was discussed both in the 'Kernforschungsanlage Juelich' and in the 'Rheinische Braunkohlenwerke' independent of each other. In 1975 these two organizations concluded a co-operation contract. (orig.) [de

  14. Nuclear energy and process heating

    International Nuclear Information System (INIS)

    Kozier, K.S.

    1999-10-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating and several industrial applications. Although only About 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven, a determined

  15. Nuclear energy and process heating

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-10-01

    Nuclear energy generated in fission reactors is a versatile commodity that can, in principle, satisfy any and all of mankind's energy needs through direct or indirect means. In addition to its dominant current use for electricity generation and, to a lesser degree, marine propulsion, nuclear energy can and has been used for process heat applications, such as space heating, industrial process heating and seawater desalination. Moreover, a wide variety of reactor designs has been employed to this end in a range of countries. From this spectrum of experience, two design approaches emerge for nuclear process heating (NPH): extracting a portion of the thermal energy from a nuclear power plant (NPP) (i.e., creating a combined heat and power, or CHP, plant) and transporting it to the user, or deploying dedicated nuclear heating plants (NHPs) in generally closer proximity to the thermal load. While the former approach is the basis for much of the current NPH experience, considerable recent interest exists for the latter, typically involving small, innovative reactor plants with enhanced and passive safety features. The high emphasis on inherent nuclear safety characteristics in these reactor designs reflects the need to avoid any requirement for evacuation of the public in the event of an accident, and the desire for sustained operation and investment protection at minimum cost. Since roughly 67% of mankind's primary energy usage is not in the form of electricity, a vast potential market for NPH systems exists, particularly at the low-to-moderate end-use temperatures required for residential space heating and several industrial applications. Although only About 0.5% of global nuclear energy production is presently used for NPH applications, an expanded role in the 21st century seems inevitable, in part, as a measure to reduce greenhouse gas emissions and improve air quality. While the technical aspects of many NPH applications are considered to be well proven, a

  16. Determination of temperature measurements uncertainties of the heat transport primary system of Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Pomerantz, Marcelo E.; Coutsiers, Eduardo E.; Moreno, Carlos A.

    1999-01-01

    In this work, the systematic errors in temperature measurements in inlet and outlet headers of HTPS coolant channels of Embalse nuclear power plant are evaluated. These uncertainties are necessary for a later evaluation of the channel power maps transferred to the coolant. The power maps calculated in this way are used to compare power distributions using neutronic codes. Therefore, a methodology to correct systematic errors of temperature in outlet feeders and inlet headers is developed in this work. (author)

  17. Analysis of heat and mass transport processes near an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Keller, C.

    1990-01-01

    A review has been performed of the models and experimental plans for evaluation of the spent fuel canister environment in a nuclear repository, e.g., the planned Yucca Mountain facilities. Special emphasis was placed on the relevance of the models and experiments to the 100 to 10,000 year prediction. The question was addressed whether one could justify testing in materials other than Yucca Mountain rock and obtain results in a relatively short time which would be relevant to the long time in Yucca Mountain. The paper discusses steam evolution in calculations and experiments, fracture models, possible measurements of relative permeability, and long time scale effects. 5 figs. (MB)

  18. Chances for nuclear district heat generation

    International Nuclear Information System (INIS)

    Winkens, H.P.

    1986-01-01

    Nuclear power plants in the FRG or other European countries so far have not been intended for heat generation, as for reasons of safety they have to be sited too far away from urban agglomerations to make heat transport competible. In addition, heat generation costs of fossil-fueled power plants have not been so much higher than those of nuclear power stations that the extra cost for heat transport over large distances could have been justified. This situation is expected to gradually change over the next decade, as the heat from fossil-fueled power stations will become more expensive, as a result of this heat capacity being more and more used for medium-load and peak-load supply only, and with more efficient heat distribution systems becoming available in the near future. (orig.) [de

  19. Nuclear materials transport worldwide

    International Nuclear Information System (INIS)

    Stellpflug, J.

    1987-01-01

    This Greenpeace report shows: nuclear materials transport is an extremely hazardous business. There is no safe protection against accidents, kidnapping, or sabotage. Any moment of a day, at any place, a nuclear transport accident may bring the world to disaster, releasing plutonium or radioactive fission products to the environment. Such an event is not less probable than the MCA at Chernobyl. The author of the book in hand follows the secret track of radioactive materials around the world, from uranium mines to the nuclear power plants, from reprocessing facilities to the waste repositories. He explores the routes of transport and the risks involved, he gives the names of transport firms and discloses incidents and carelessness, tells about damaged waste drums and plutonium that 'disappeared'. He also tells about worldwide, organised resistance to such nuclear transports, explaining the Greenpeace missions on the open sea, or the 'day X' operation at the Gorleben site, informing the reader about protests and actions for a world freed from the threat of nuclear energy. (orig./HP) [de

  20. Heat recovery from nuclear power plants

    International Nuclear Information System (INIS)

    Safa, H.

    2012-01-01

    The thermodynamic efficiency of a standard Nuclear Power Plant (NPP) is around 33%. Therefore, about two third of the heat generated by the nuclear fuel is literally wasted in the environment. Given the fact that the steam coming out from the high pressure turbine is superheated, it could be advantageously used for non electrical applications, particularly for district heating. Considering the technological improvements achieved these last years in heat piping insulation, it is now perfectly feasible to envisage heat transport over quite long distances, exceeding 200 km, with affordable losses. Therefore, it could be energetically wise to revise the modifications required on present reactors to perform heat extraction without impeding the NPP operation. In this paper, the case of a French reactor is studied showing that a large fraction of the wasted nuclear heat can be actually recovered and transported to be injected in the heat distribution network of a large city. Some technical and economical aspects of nuclear district heating application are also discussed. (author)

  1. Heat transport through atomic contacts.

    Science.gov (United States)

    Mosso, Nico; Drechsler, Ute; Menges, Fabian; Nirmalraj, Peter; Karg, Siegfried; Riel, Heike; Gotsmann, Bernd

    2017-05-01

    Heat transport and dissipation at the nanoscale severely limit the scaling of high-performance electronic devices and circuits. Metallic atomic junctions serve as model systems to probe electrical and thermal transport down to the atomic level as well as quantum effects that occur in one-dimensional (1D) systems. Whereas charge transport in atomic junctions has been studied intensively in the past two decades, heat transport remains poorly characterized because it requires the combination of a high sensitivity to small heat fluxes and the formation of stable atomic contacts. Here we report heat-transfer measurements through atomic junctions and analyse the thermal conductance of single-atom gold contacts at room temperature. Simultaneous measurements of charge and heat transport reveal the proportionality of electrical and thermal conductance, quantized with the respective conductance quanta. This constitutes a verification of the Wiedemann-Franz law at the atomic scale.

  2. Nuclear power for district heating

    International Nuclear Information System (INIS)

    Lyon, R.B.; Sochaski, R.O.

    1975-09-01

    Current district heating trends are towards an increasing use of electricity. This report concerns the evaluation of an alternative means of energy supply - the direct use of thermal energy from CANDU nuclear stations. The energy would be transmitted via a hot fluid in a pipeline over distances of up to 40 km. Advantages of this approach include a high utilization of primary energy, with a consequent reduction in installed capacity, and load flattening due to inherent energy storage capacity and transport delays. Disadvantages include the low load factors for district heating, the high cost of the distribution systems and the necessity for large-scale operation for economic viability. This requirement for large-scale operation from the beginning could cause difficulty in the implementation of the first system. Various approaches have been analysed and costed for a specific application - the supply of energy to a district heating load centre in Toronto from the location of the Pickering reactor station about 40 km away. (author)

  3. Valve stem packing seal test results for primary heat transport system conditions in Canadian nuclear generating stations

    International Nuclear Information System (INIS)

    Dixon, D.F.; Farrell, J.M.; Coutinho, R.F.

    1978-06-01

    Valve stem packing tests were done to obtain performance data on packing already in CANDU-PHW reactor service and on alternative packings. Most of the tests were replicated. Results are presented for ten packings tested under two stem cycle modes; leakage, packing consolidation and packing friction were the main responses. Packing tests were performed with water at close to CANDU-PHW reactor primary heat transport (PHT) system conditions (288 deg C and 10 MPa), but without ionizing radiation. The test rigs had rising, rotating stems. Stuffing box dimensions were typical of a standard Velan valve; packings were spring loaded to control applied packing stress

  4. Transportation of nuclear materials

    International Nuclear Information System (INIS)

    Brobst, W.A.

    1977-01-01

    Twenty years of almost accident-free transport of nuclear materials is pointed to as evidence of a fundamentally correct approach to the problems involved. The increased volume and new technical problems in the future will require extension of these good practices in both regulations and packaging. The general principles of safety in the transport of radioactive materials are discussed first, followed by the transport of spent fuel and of radioactive waste. The security and physical protection of nuclear shipments is then treated. In discussing future problems, the question of public understanding and acceptance is taken first, thereafter transport safeguards and the technical bases for the safety regulations. There is also said to be a need for a new technology for spent fuel casks, while a re-examination of the IAEA transport standards for radiation doses is recommended. The IAEA regulations regarding quality assurance are said to be incomplete, and more information is required on correlations between engineering analysis, scale model testing and full scale crash testing. Transport stresses on contents need to be considered while administrative controls have been neglected. (JIW)

  5. Temperature in the Primary Heat Transport Pump Bearing of the Nuclear Power Plant 'Embalse Rio Tercero' in view of the Cutting of the Service Water

    International Nuclear Information System (INIS)

    Raffo, J.L

    2001-01-01

    This study contains the analysis of the Primary Heat Transport Pump Bearing of the Nuclear Power Plant 'Embalse Rio Tercero', Cordoba, Argentine, in view of the cutting of the Service Water refrigeration which cools the Gland Seal System.This takes two ways: One is the study of the temperature rise of the water that cools the carbon bearing and the time involved.This is supported upon manuals and drawings.The other, on the temperature distribution in different operating conditions.This has been done by the simulation of the normal and transient conditions in the software COSMOS/M

  6. Heat pump augmentation of nuclear process heat

    International Nuclear Information System (INIS)

    Koutz, S.L.

    1986-01-01

    A system is described for increasing the temperature of a working fluid heated by a nuclear reactor. The system consists of: a high temperature gas cooled nuclear reactor having a core and a primary cooling loop through which a coolant is circulated so as to undergo an increase in temperature, a closed secondary loop having a working fluid therein, the cooling and secondary loops having cooperative association with an intermediate heat exchanger adapted to effect transfer of heat from the coolant to the working fluid as the working fluid passes through the intermediate heat exchanger, a heat pump connected in the secondary loop and including a turbine and a compressor through which the working fluid passes so that the working fluid undergoes an increase in temperature as it passes through the compressor, a process loop including a process chamber adapted to receive a process fluid therein, the process chamber being connected in circuit with the secondary loop so as to receive the working fluid from the compressor and transfer heat from the working fluid to the process fluid, a heat exchanger for heating the working fluid connected to the process loop for receiving heat therefrom and for transferring heat to the secondary loop prior to the working fluid passing through the compressor, the secondary loop being operative to pass the working fluid from the process chamber to the turbine so as to effect driving relation thereof, a steam generator operatively associated with the secondary loop so as to receive the working fluid from the turbine, and a steam loop having a feedwater supply and connected in circuit with the steam generator so that feedwater passing through the steam loop is heated by the steam generator, the steam loop being connected in circuit with the process chamber and adapted to pass steam to the process chamber with the process fluid

  7. Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

    Directory of Open Access Journals (Sweden)

    Avinash J. Gaikwad

    2008-01-01

    Full Text Available Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs, like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs is resorted to mitigate consequences of station blackout (SBO. In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR system is also incorporated to condense the steam generated in the boilers during a SBO. The sustainability of natural circulation in the various heat transport systems (i.e., primary heat transport (PHT, SGs, and PDHRs under station blackout depends on the corresponding system's coolant inventories and the coolant circuit configurations (i.e., parallel paths and interconnections. On the primary side, the interconnection between the two primary loops plays an important role to sustain the natural circulation heat removal. On the secondary side, the steam lines interconnections and the initial inventory in the SGs prior to cooldown, that is, hooking up of the PDHRs are very important. This paper attempts to open up discussions on the concept and the core issues associated with passive systems which can provide continued heat sink during such accident scenarios. The discussions would include the criteria for design, and performance of such concepts already implemented and proposes schemes to be implemented in the proposed 700 MWe IPHWR. The designer feedbacks generated, and critical examination of performance analysis results for the added passive system to the existing generation II & III reactors will help ascertaining that these safety systems/inventories in fact perform in sustaining decay heat removal and augmenting safety.

  8. Acoustically enhanced heat transport

    Energy Technology Data Exchange (ETDEWEB)

    Ang, Kar M.; Hung, Yew Mun; Tan, Ming K., E-mail: tan.ming.kwang@monash.edu [School of Engineering, Monash University Malaysia, 47500 Bandar Sunway, Selangor (Malaysia); Yeo, Leslie Y. [Micro/Nanophysics Research Laboratory, RMIT University, Melbourne, VIC 3001 (Australia); Friend, James R. [Department of Mechanical and Aerospace Engineering, University of California, San Diego, California 92093 (United States)

    2016-01-15

    We investigate the enhancement of heat transfer in the nucleate boiling regime by inducing high frequency acoustic waves (f ∼ 10{sup 6} Hz) on the heated surface. In the experiments, liquid droplets (deionized water) are dispensed directly onto a heated, vibrating substrate. At lower vibration amplitudes (ξ{sub s} ∼ 10{sup −9} m), the improved heat transfer is mainly due to the detachment of vapor bubbles from the heated surface and the induced thermal mixing. Upon increasing the vibration amplitude (ξ{sub s} ∼ 10{sup −8} m), the heat transfer becomes more substantial due to the rapid bursting of vapor bubbles happening at the liquid-air interface as a consequence of capillary waves travelling in the thin liquid film between the vapor bubble and the air. Further increases then lead to rapid atomization that continues to enhance the heat transfer. An acoustic wave displacement amplitude on the order of 10{sup −8} m with 10{sup 6} Hz order frequencies is observed to produce an improvement of up to 50% reduction in the surface temperature over the case without acoustic excitation.

  9. Acoustically enhanced heat transport

    Science.gov (United States)

    Ang, Kar M.; Yeo, Leslie Y.; Friend, James R.; Hung, Yew Mun; Tan, Ming K.

    2016-01-01

    We investigate the enhancement of heat transfer in the nucleate boiling regime by inducing high frequency acoustic waves (f ˜ 106 Hz) on the heated surface. In the experiments, liquid droplets (deionized water) are dispensed directly onto a heated, vibrating substrate. At lower vibration amplitudes (ξs ˜ 10-9 m), the improved heat transfer is mainly due to the detachment of vapor bubbles from the heated surface and the induced thermal mixing. Upon increasing the vibration amplitude (ξs ˜ 10-8 m), the heat transfer becomes more substantial due to the rapid bursting of vapor bubbles happening at the liquid-air interface as a consequence of capillary waves travelling in the thin liquid film between the vapor bubble and the air. Further increases then lead to rapid atomization that continues to enhance the heat transfer. An acoustic wave displacement amplitude on the order of 10-8 m with 106 Hz order frequencies is observed to produce an improvement of up to 50% reduction in the surface temperature over the case without acoustic excitation.

  10. Gasification with nuclear reactor heat

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1977-01-01

    The energy-political ultimate aims for the introduction of nuclear coal gasification and the present state of technology concerning the HTR reactor, concerning gasification and heat exchanging components are outlined. Presented on the plans a) for hydro-gasification of lignite and for steam gasification of pit coal for the production of synthetic natural gas, and b) for the introduction of a nuclear heat system. The safety and environmental problems to be expected are portrayed. The main points of development, the planned prototype plant and the schedule of the project Pototype plant Nuclear Process heat (PNP) are specified. In a market and economic viability study of nuclear coal gasification, the application potential of SNG, the possible construction programme for the FRG, as well as costs and rentability of SNG production are estimated. (GG) [de

  11. Impact of thermal conductivity models on the coupling of heat transport, oxygen diffusion, and deformation in (U, Pu)O nuclear fuel elements

    Science.gov (United States)

    Mihaila, Bogdan; Stan, Marius; Crapps, Justin; Yun, Di

    2013-02-01

    We study the coupled thermal transport, oxygen diffusion, and thermal expansion in a generic nuclear fuel rod consisting of a (U) fuel pellet separated by a helium gap from zircaloy cladding. Steady-state and time-dependent finite-element simulations with a variety of initial- and boundary-value conditions are used to study the effect of the Pu content, y, and deviation from stoichiometry, x, on the temperature and deformation profiles in this fuel element. We find that the equilibrium radial temperature and deformation profiles are most sensitive to x at small values of y. For larger values of y, the effects of oxygen and Pu content are equally important. Following a change in the heat-generation rate, the centerline temperature, the radial deformation of the fuel pellet, and the centerline deviation from stoichiometry track each other closely in (U,Pu)O, as the characteristic time scales of the heat transport and oxygen diffusion are similar. This result is different from the situation observed in the case of UO fuels.

  12. Safety study on nuclear heat utilization system - accident delineation and assessment on nuclear steelmaking pilot plant

    International Nuclear Information System (INIS)

    Yoshida, T.; Mizuno, M.; Tsuruoka, K.

    1982-01-01

    This paper presents accident delineation and assessment on a nuclear steelmaking pilot plant as an example of nuclear heat utilization systems. The reactor thermal energy from VHTR is transported to externally located chemical process plant employing helium-heated steam reformer by an intermediate heat transport loop. This paper on the nuclear steelmaking pilot plant will describe (1) system transients under accident conditions, (2) impact of explosion and fire on the nuclear reactor and the public and (3) radiation exposure on the public. The results presented in this paper will contribute considerably to understanding safety features of nuclear heat utilization system that employs the intermediate heat transport loop and the helium-heated steam reformer

  13. SECURE nuclear district heating plant

    International Nuclear Information System (INIS)

    Nilsson; Hannus, M.

    1978-01-01

    The role foreseen for the SECURE (Safe Environmentally Clean Urban REactor) nuclear district heating plant is to provide the baseload heating needs of primarily the larger and medium size urban centers that are outside the range of waste heat supply from conventional nuclear power stations. The rationale of the SECURE concept is that the simplicity in design and the inherent safety advantages due to the use of low temperatures and pressures should make such reactors economically feasible in much smaller unit sizes than nuclear power reactors and should make their urban location possible. It is felt that the present design should be safe enough to make urban underground location possible without restriction according to any criteria based on actual risk evaluation. From the environmental point of view, this is a municipal heat supply plant with negligible pollution. Waste heat is negligible, gaseous radioactivity release is negligible, and there is no liquid radwaste release. Economic comparisons show that the SECURE plant is competitive with current fossil-fueled alternatives. Expected future increase in energy raw material prices will lead to additional energy cost advantages to the SECURE plant

  14. Transportation of spent nuclear fuels

    International Nuclear Information System (INIS)

    Meguro, Toshiichi

    1976-01-01

    The spent nuclear fuel taken out of reactors is cooled in the cooling pool in each power station for a definite time, then transported to a reprocessing plant. At present, there is no reprocessing plant in Japan, therefore the spent nuclear fuel is shipped abroad. In this paper, the experiences and the present situation in Japan are described on the transport of the spent nuclear fuel from light water reactors, centering around the works in Tsuruga Power Station, Japan Atomic Power Co. The spent nuclear fuel in Tsuruga Power Station was first transported in Apr. 1973, and since then, about 36 tons were shipped to Britain by 5 times of transport. The reprocessing plant in Japan is expected to start operation in Apr. 1977, accordingly the spent nuclear fuel used for the trial will be transported in Japan in the latter half of this year. Among the permission and approval required for the transport of spent nuclear fuel, the acquisition of the certificate for transport casks and the approval of land and sea transports are main tasks. The relevant laws are the law concerning the regulations of nuclear raw material, nuclear fuel and reactors and the law concerning the safety of ships. The casks used in Tsuruga Power Station and EXL III type, and the charging of spent nuclear fuel, the decontamination of the casks, the leak test, land transport with a self-running vehicle, loading on board an exclusive carrier and sea transport are briefly explained. The casks and the ship for domestic transport are being prepared. (Kato, I.)

  15. Nuclear Transport of Yeast Proteasomes

    Directory of Open Access Journals (Sweden)

    Cordula Enenkel

    2014-10-01

    Full Text Available Proteasomes are conserved protease complexes enriched in the nuclei of dividing yeast cells, a major site for protein degradation. If yeast cells do not proliferate and transit to quiescence, metabolic changes result in the dissociation of proteasomes into proteolytic core and regulatory complexes and their sequestration into motile cytosolic proteasome storage granuli. These granuli rapidly clear with the resumption of growth, releasing the stored proteasomes, which relocalize back to the nucleus to promote cell cycle progression. Here, I report on three models of how proteasomes are transported from the cytoplasm into the nucleus of yeast cells. The first model applies for dividing yeast and is based on the canonical pathway using classical nuclear localization sequences of proteasomal subcomplexes and the classical import receptor importin/karyopherin αβ. The second model applies for quiescent yeast cells, which resume growth and use Blm10, a HEAT-like repeat protein structurally related to karyopherin β, for nuclear import of proteasome core particles. In the third model, the fully-assembled proteasome is imported into the nucleus. Our still marginal knowledge about proteasome dynamics will inspire the discussion on how protein degradation by proteasomes may be regulated in different cellular compartments of dividing and quiescent eukaryotic cells.

  16. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    International Nuclear Information System (INIS)

    Iwamoto, Y.; Ogawa, T.

    2016-01-01

    The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials

  17. High temperature heat exchange: nuclear process heat applications

    International Nuclear Information System (INIS)

    Vrable, D.L.

    1980-09-01

    The unique element of the HTGR system is the high-temperature operation and the need for heat exchanger equipment to transfer nuclear heat from the reactor to the process application. This paper discusses the potential applications of the HTGR in both synthetic fuel production and nuclear steel making and presents the design considerations for the high-temperature heat exchanger equipment

  18. Nuclear transport - The regulatory dimension

    International Nuclear Information System (INIS)

    Green, L.

    2002-01-01

    The benefits that the peaceful applications of nuclear energy have brought to society are due in no small part to industry's capacity to transport radioactive materials safely, efficiently and reliably. The nuclear transport industry has a vital role in realising a fundamental objective of the International Atomic Energy Agency (IAEA) as stated in its statute to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world. The context in which transports currently take place is complex, and rapidly changing. In many respects transport is being viewed as an integral market issue and not a subsidiary concern. The availability of carriers drives routing decisions and changes in material flows necessitate new approaches to packaging and transport scenarios. Pressures on the transport sector are not without serious consequences; they can cause delays and in some cases cancellation of planned movements. Complex routings and the necessary use of chartered carriers can push up costs and work against cost efficiency. Since the events of 11 September 2001 the security of nuclear transports has contributed an added dimension to how transports take place. Transports of radioactive material have an outstanding safety record, indeed the transport of such materials could be regarded as a model for the transport of other classes of dangerous goods. This safety record is achieved by two inter-related factors. It is due primarily to well founded regulations developed by such key intergovernmental organisations as the IAEA, with the essential contributions of the member states who participate in the implementation of regulations and the review process. It is due also to the professionalism of those in the industry. There is a necessary synergy between the two - between the regulators whose task it is to make and to enforce the rules for safe, efficient and reliable transport and those whose job it is to transport within the rules. It

  19. Nuclear materials transport in France

    International Nuclear Information System (INIS)

    Korycanek, J.

    1990-01-01

    About 1.5 million tons of uranium ore, 8000 tons of uranium concentrate, 1000 tons of UF 6 , 340 spent fuel containers, and 30 000 m 3 of nuclear wastes are transported annually by trucks, trains and ships in France. Annual costs of this transportation amount to 500-600 million FRF, and about 200 employees are engaged in this activity. Transportation of spent fuel to the La Hague and Marcoule fuel reprocessing plants, and the transport of plutonium are dealt with in detail. (Z.M.). 5 figs., 1 ref

  20. Supply of Prague with heat from a nuclear heat source

    International Nuclear Information System (INIS)

    Poul, F.

    1976-01-01

    The proposals are discussed of supplying Prague, the Czechoslovak Capital, with nuclear reactor-generated heat energy. The proposals meet the requirements of the general urban plan of development. The first nuclear heating plant is to be sited in the Kojetice locality, in the northern Prague suburb. It will be commissioned by 1984 and 1985. It is estimated that the maximum heat output in form of hot water will be 821 MW. By 1995 the construction of the second nuclear heating plant should be started southeast or east of Prague. The connection of these two nuclear plants to the hot water mains together with other conventional heating plants will secure the heat supply for Prague and its new housing estates and industrial works. (Oy)

  1. Nuclear heat for industrial purposes and district heating

    International Nuclear Information System (INIS)

    1974-01-01

    Studies on the various possibilities for the application of heat from nuclear reactors in the form of district heat or process steam for industrial purposes had been made long before the present energy crisis. Although these studies have indicated technical feasibility and economical justification of such utilization, the availability of relatively cheap oil and difficulties in locating a nuclear heat source inside industrial areas did not stimulate much further development. Since the increase of oil prices, the interest in nuclear heat application is reawakened, and a number of new potential areas have been identified. It now seems generally recognized that the heat from nuclear reactors should play an important role in primary energy supply, not only for electricity production but also as direct heat. At present three broad areas of nuclear heat application are identified: Direct heat utilization in industrial processing requiring a temperature above 800 deg. C; Process steam utilization in various industries, requiring a temperature mainly in the range of 200-300 deg. C; Low temperature and waste heat utilization from nuclear power plants for desalination of sea water and district heating. Such classification is mainly related to the type and characteristics of the heat source or nuclear reactor which could be used for a particular application. Modified high temperature reactor types (HTR) are the candidates for direct heat application, while the LWR reactors can satisfy most of the demands for process steam. Production of waste heat is a characteristic of all thermal power plants, and its utilization is a major challenge in the field of power production

  2. Inspection of nuclear fuel transport in Spain

    International Nuclear Information System (INIS)

    Lobo Mendez, J.

    1977-01-01

    The experience acquired in inspecting nuclear fuel shipments carried out in Spain will serve as a basis for establishing the regulations wich must be adhered to for future transports, as the transport of nuclear fuels in Spain will increase considerably within the next years as a result of the Spanish nuclear program. The experience acquired in nuclear fuel transport inspection is described. (author) [es

  3. Effect of Coolant Inventories and Parallel Loop Interconnections on the Natural Circulation in Various Heat Transport Systems of a Nuclear Power Plant during Station Blackout

    OpenAIRE

    Gaikwad, Avinash J.; Vijayan, P. K.; Bhartya, Sharad; Iyer, Kannan; Kumar, Rajesh; Contractor, A. D.; Lele, H. G.; Vhora, S. F.; Maurya, A. K.; Ghosh, A. K.; Kushwaha, H. S.

    2008-01-01

    Provision of passive means to reactor core decay heat removal enhances the nuclear power plant (NPP) safety and availability. In the earlier Indian pressurised heavy water reactors (IPHWRs), like the 220 MWe and the 540 MWe, crash cooldown from the steam generators (SGs) is resorted to mitigate consequences of station blackout (SBO). In the 700 MWe PHWR currently being designed an additional passive decay heat removal (PDHR) system is also incorporated to condense the steam generated in the b...

  4. Survey of high-temperature nuclear heat application

    International Nuclear Information System (INIS)

    Kirch, N.; Schaefer, M.

    1984-01-01

    Nuclear heat application at high temperatures can be divided into two areas - use of high-temperature steam up to 550 deg. C and use of high-temperature helium up to about 950 deg. C. Techniques of high-temperature steam and heat production and application are being developed in several IAEA Member States. In all these countries the use of steam for other than electricity production is still in a project definition phase. Plans are being discussed about using steam in chemical industries, oil refineries and for new synfuel producing plants. The use of nuclear generated steam for oil recovery from sands and shale is also being considered. High-temperature nuclear process heat production gives new possibilities for the application of nuclear energy - hard coals, lignites, heavy oils, fuels with problems concerning transport, handling and pollution can be converted into gaseous or liquid energy carriers with no loss of their energy contents. The main methods for this conversion are hydrogasification with hydrogen generated by nuclear heated steam reformers and steam gasification. These techniques will allow countries with large coal resources to replace an important part of their natural gas and oil consumption. Even countries with no fossil fuels can benefit from high-temperature nuclear heat - hydrogen production by thermochemical water splitting, nuclear steel making, ammonia production and the chemical heat-pipe system are examples in this direction. (author)

  5. Transport processes: Momentum, heat and mass

    International Nuclear Information System (INIS)

    Geankoplis, C.J.

    1983-01-01

    This book discusses basic transport processes including mass transport. Topics covered are as follows: an introduction to engineering principles and units; principles of momentum transfer and overall balances; principles of momentum transfer and applications; principles of steady-state heat transfer; principles of unsteady-state heat transfer; principles of mass transfer; principles of unsteady-state and convective mass transfer

  6. Ductile fracture behaviour of primary heat transport piping material ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    1. Introduction. The design of primary heat transport piping (PHT) of nuclear reactors has to ensure that uncontrolled failure does not occur under normal, faulted or accidental service conditions. One of the most severe failure scenarios traditionally envisaged is instantaneous double-ended guillotine break (DEGB), which ...

  7. Design and safety aspects of nuclear district heating reactors

    International Nuclear Information System (INIS)

    Brogli, R.; Mathews, D.; Pelloni, S.

    1989-01-01

    Extensive studies on the rationale, the potential and the technology of nuclear district heating have been performed in Switzerland. Beside economics the safety aspects were of primary importance. Due to the high costs to transport heat the heating reactor tend to be small and therefore, minimally staffed and located close to population centers. Stringed safety rules are therefore applying. Gas cooled reactors are well suited as district heating reactors since they have due to their characteristics several inherent features, significant safety margins and a remarkable radioactivity retention potential. Some ways to mitigate the effects of water ingress and graphite corrosion are under investigation. (author). 5 refs, 3 figs

  8. TOUGH, Unsaturated Groundwater Transport and Heat Transport Simulation

    International Nuclear Information System (INIS)

    Pruess, K.A.; Cooper, C.; Osnes, J.D.

    1992-01-01

    1 - Description of program or function: A successor to the TOUGH program, TOUGH2 offers added capabilities and user features, including the flexibility to handle different fluid mixtures (water, water with tracer; water, CO 2 ; water, air; water, air with vapour pressure lowering, and water, hydrogen), facilities for processing of geometric data (computational grids), and an internal version control system to ensure referenceability of code applications. TOUGH (Transport of Unsaturated Groundwater and Heat) is a multi-dimensional numerical model for simulating the coupled transport of water, vapor, air, and heat in porous and fractured media. The program provides options for specifying injection or withdrawal of heat and fluids. Although primarily designed for studies of high-level nuclear waste isolation in partially saturated geological media, it should also be useful for a wider range of problems in heat and moisture transfer, and in the drying of porous materials. For example, geothermal reservoir simulation problems can be handled simply by setting the air mass function equal to zero on input. The TOUGH simulator was developed for problems involving strongly heat-driven flow. To describe these phenomena a multi-phase approach to fluid and heat flow is used, which fully accounts for the movement of gaseous and liquid phases, their transport of latent transitions between liquid and vapor. TOUGH takes account of fluid flow in both liquid and gaseous phases occurring under pressure, viscous, and gravity forces according to Darcy's law. Interference between the phases is represented by means of relative permeability functions. The code handles binary, but not Knudsen, diffusion in the gas phase and capillary and phase absorption effects for the liquid phase. Heat transport occurs by means of conduction with thermal conductivity dependent on water saturation, convection, and binary diffusion, which includes both sensible and latent heat. 2 - Method of solution: All

  9. Perspective on transporting nuclear materials

    International Nuclear Information System (INIS)

    Wymer, R.G.

    1975-01-01

    An evaluation is made of the material flow to be expected up to the year 2000 to and from the various steps in the nuclear cycle. These include the reactors, reprocessing plants, enrichment plants, U mills, U conversion plants, and fuel fabrication plants. A somewhat more-detailed discussion is given of the safety engineering that goes into the design of containers and packages and the selection of the mode of transportation. The relationship of shipping to siting and transportation accidents is considered briefly

  10. THERMOS, district central heating nuclear reactors

    International Nuclear Information System (INIS)

    Patarin, L.

    1981-02-01

    In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr

  11. Transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    1980-01-01

    In response to public interest in the transport by rail through London of containers of irradiated fuel elements on their way from nuclear power stations to Windscale, the Central Electricity Generating Board and British Rail held three information meetings in London in January 1980. One meeting was for representatives of London Borough Councils and Members of Parliament with a known interest in the subject, and the others were for press, radio and television journalists. This booklet contains the main points made by the principal speakers from the CEGB and BR. (The points covered include: brief description of the fuel cycle; effect of the fission process in producing plutonium and fission products in the fuel element; fuel transport; the fuel flasks; protection against accidents; experience of transporting fuel). (U.K.)

  12. A nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Bancroft, A.R.; Fenton, N.

    1989-07-01

    Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

  13. Low-temperature nuclear heat applications: Nuclear power plants for district heating

    International Nuclear Information System (INIS)

    1987-08-01

    The IAEA reflected the needs of its Member States for the exchange of information in the field of nuclear heat application already in the late 1970s. In the early 1980s, some Member States showed their interest in the use of heat from electricity producing nuclear power plants and in the development of nuclear heating plants. Accordingly, a technical committee meeting with a workshop was organized in 1983 to review the status of nuclear heat application which confirmed both the progress made in this field and the renewed interest of Member States in an active exchange of information about this subject. In 1985 an Advisory Group summarized the Potential of Low-Temperature Nuclear Heat Application; the relevant Technical Document reviewing the situation in the IAEA's Member States was issued in 1986 (IAEA-TECDOC-397). Programme plans were made for 1986-88 and the IAEA was asked to promote the exchange of information, with specific emphasis on the design criteria, operating experience, safety requirements and specifications for heat-only reactors, co-generation plants and power plants adapted for heat application. Because of a growing interest of the IAEA's Member States about nuclear heat employment in the district heating domaine, an Advisory Group meeting was organized by the IAEA on ''Low-Temperature Nuclear Heat Application: Nuclear Power Plants for District Heating'' in Prague, Czechoslovakia in June 1986. The information gained up to 1986 and discussed during this meeting is embodied in the present Technical Document. 22 figs, 11 tabs

  14. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

    A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

  15. French nuclear power plants for heat generation

    International Nuclear Information System (INIS)

    Girard, Y.

    1984-01-01

    The considerable importance that France attributes to nuclear energy is well known even though as a result of the economic crisis and the energy savings it is possible to observe a certain downward trend in the rate at which new power plants are being started up. In July 1983, a symbolic turning-point was reached - at more than 10 thousand million kW.h nuclear power accounted, for the first time, for more than 50% of the total amount of electricity generated, or approx. 80% of the total electricity output of thermal origin. On the other hand, the direct contribution - excluding the use of electricity - of nuclear energy to the heat market in France remains virtually nil. The first part of this paper discusses the prospects and realities of the application, at low and intermediate temperatures, of nuclear heat in France, while the second part describes the French nuclear projects best suited to the heat market (excluding high temperatures). (author)

  16. The prospects for nuclear heating in Hungary

    International Nuclear Information System (INIS)

    Papp, I.; Lynch, G.F.

    1989-09-01

    In assessing alternative nuclear heat sources, a joint study was undertaken between Canada and Hungary to determine the feasibility of using the SLOWPOKE Energy System that has recently been developed. The SLOWPOKE Energy System is a benign nuclear heat source designed to supply 10 thermal megawatts in the form of hot water for local heating systems in buildings and institutions. It uses a combination of inherent safety features, including natural convection circulation and negative reactivity coefficients, and engineered features to ensure an extremely safe system. A SLOWPOKE demonstration heating reactor has been constructed in Canada. The unit started operation in July 1987 and is currently undergoing an extensive test program. Since the nuclear heat source is small, operates at atmospheric pressure, and produces hot water below 100 deg. C, the complex high-pressure, and high-temperature systems essential for electricity production are eliminated. As a result, the nuclear heat source can be located close to the load and will require a minimum of operator attention. In this way, a SLOWPOKE Energy System can be considered much like the oil- or natural gas fired furnace it is designed to replace. The extensive use of hot water district heating systems in Hungary offers the opportunity to exploit such simple nuclear systems as base load heat sources without an extensive retrofit of the existing systems. In addition, the studies have concluded that there are many economically attractive sites for 10 MW SLOWPOKE Energy Systems within the existing networks. The low capital investment requirements, coupled with a high degree of localization, even for the first unit, are seen as additional factors that facilitate the transfer of the technology to Hungary. Simple nuclear heat sources, such as the SLOWPOKE Energy System, when applied to the Hungarian district heating systems, offer the prospects of a significant reduction in the dependence on imported fossil fuels in the

  17. Ion heat transport studies in JET

    DEFF Research Database (Denmark)

    Mantica, P; Angioni, C; Baiocchi, B

    2011-01-01

    Detailed experimental studies of ion heat transport have been carried out in JET exploiting the upgrade of active charge exchange spectroscopy and the availability of multi-frequency ion cyclotron resonance heating with 3He minority. The determination of ion temperature gradient (ITG) threshold...

  18. Oil shales and the nuclear process heat

    International Nuclear Information System (INIS)

    Scarpinella, C.A.

    1974-01-01

    Two of the primary energy sources most dited as alternatives to the traditional fossil fuels are oil shales and nuclear energy. Several proposed processes for the extraction and utilization of oil and gas from shale are given. Possible efficient ways in which nuclear heat may be used in these processes are discussed [pt

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Engineering and economic aspects of centalized heating from nuclear boilers

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Baturov, B.B.; Korytnikov, V.P.; Koryakin, Yu.I.; Chernyaev, V.A.; Kovylyanskij, Ya.A.; Galaktionov, I.V.

    1979-01-01

    Some engineering and economic aspects for deployment of centralized nuclear boilers (NB) in the USSR are considered. Engineering, maintenance and economic features of NB as compared to organic-fuelled boilers and nuclear thermal power plants are discussed. Among major factors governing economic efficiency of NB underlined are oraganic fuel costs, reactor unit power, location relative to heat-consuming centres and capacity factor. It is concluded that NB can be economical for heating large consumers (more than 1500 G kal/hr). At the periphery NB can be competitive already at reactor unit power of several MWth. The development of HTGR type reactor-based nuclear-chemical boilers and lines for heat transport in a chemically bound state (e.g., CH 4 → H 2 +CO 2 +CO → CH 4 ) opens the way for a substantial breakthrow in the centralized NB efficiency

  1. Equipment transporter for nuclear steam generator

    International Nuclear Information System (INIS)

    Hayes, L.R.

    1987-01-01

    A transporter is described for use in a steam generator of a nuclear power installation. The generator is essentially a heat exchanger having a vertically extended shell. Across the lower portion extends a horizontal tube sheet having an upper surface which supports a bundle of vertically extending tubes forming a limited annular space with the inside of the shell wall and the upper surface. An opening of limited dimensions through the shell wall gains manual access to the limited annular space. The transporter has means for locating and removing solid debris from the upper surface of the tube sheet in the annular space and has a means for assembly and disassembly of the transporter so that it may be manually passed through the shell opening to and from a position on the upper surface of the tube sheet in the annular space. The transporter includes: a body; at least three wheels mounted on the body for engaging the upper surface of the tube sheet; a first motor mounted on the body drivingly connected to the wheels for moving the transporter along the upper surface of the tube sheet in the annular space; a remotely operated means on the body for locating solid debris on the upper surface of the tube sheet; and means for securing and removing solid debris on the upper surface of the tube sheet located by the means for locating

  2. The sea transport of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Miller, M.L.

    1995-01-01

    The paper describes the development of a transport system dedicated to the sea transport of irradiated nuclear fuel. It reviews the background to why shipments were required and the establishment of a specialist shipping company, Pacific Nuclear Transport Limited. A description of the ships, flasks and other equipment utilized is provided, together with details of key procedures implemented to ensure safety and customer satisfaction

  3. Transport and nuclear waste disposal

    International Nuclear Information System (INIS)

    Wild, E.

    1999-01-01

    The author assesses both past and future of nuclear waste disposal in Germany. The failure of the disposal concept is, he believes, mainly the fault of the Federal Government. On the basis of the Nuclear Energy Act, the government is obliged to ensure that ultimate-storage sites are established and operated. Up to the present, however, the government has failed - apart from the episode in Asse and Morsleben and espite existing feasible proposals in Konrad and Gorleben - to achieve this objective. This negative development is particularly evident from the projects which have had to be prematurely abandoned. The costs of such 'investment follies' meanwhile amount to several billion DM. At least 92% of the capacity in the intermediate-storage sites are at present unused. Following the closure of the ultimate-storage site in Morsleben, action must be taken to change over to long-term intermediate-storage of operational waste. The government has extensive intermediate-storage capacity at the intermediate-storage site Nord in Greifswald. There, the wate originally planned for storage in Morsleben could be intermediately stored at ERAM-rates. Nuclear waste transportation, too, could long ago have been resumed, in the author's view. For the purpose of improving the transport organisation, a new company was founded which represents exclusively the interests of the reprocessing firms at the nuclear power stations. The author's conclusion: The EVU have done their homework properly and implemented all necessary measures in order to be able to resume transport of fuel elements as soon as possible. The generating station operators favour a solution based upon agreement with the Federal Government. The EVU have already declared their willingness - in the event of unanimous agreement - to set up intermediate-storage sites near the power stations. The ponds in the generating stations, however, are unsuitable for use as intermediate-storage areas. If intermediate-storage areas for

  4. Spent nuclear fuel transport problems

    International Nuclear Information System (INIS)

    Kondrat'ev, A.N.; Kosarev, Yu.A.; Yulikov, E.I.

    1977-01-01

    The paper considers the problems of shipping spent fuel from nuclear power stations to reprocessing plants and also the principal ways of solving these problems with a view to achieving maximum economy and safety in transport. The increase in the number of nuclear power plants in the USSR will entail an intensification of spent-fuel shipments. Higher burnup and the need to reduce cooling time call for heavier and more complex shipping containers. The problem of shipping spent fuel should be tackled comprehensively, bearing in mind the requirements of safety and economy. One solution to these problems is to develop rational and cheap designs of such containers. In addition, the world-wide trend towards more thorough protection of the environment against pollution and of the health of the population requires the devotion of constant attention to improving the reliability and safety of shipments. The paper considers the prospects for nuclear power development in the USSR and in other member countries of the CMEA (1976-1980), the composition and design of some Soviet packaging assemblies, the appropriate cooling time for spent fuel from thermal reactor power stations, procedures for reducing fuel-shipping costs, some methodological problems of container calculation and design, and finally problems of testing and checking containers on test rigs. (author)

  5. Hybrid district heating system with heat supply from nuclear source

    International Nuclear Information System (INIS)

    Havelka, Z.; Petrovsky, I.

    1987-01-01

    Several designs are described of heat supply from large remote power sources (e.g., WWER-1000 nuclear power plants with a 1000 MW turbine) to localities where mainly steam distribution networks have been built but only some or none networks for hot water distribution. The benefits of the designs stem from the fact that they do not require the conversion of the local steam distribution system to a hot water system. They are based on heat supply from the nuclear power plant to the consumer area in hot water of a temperature of 150 degC to 200 degC. Part of the hot water heat will be used for the production of low-pressure steam which will be compressed using heat pumps (steam compressors) to achieve the desired steam distribution network specifications. Water of lower temperature can be used in the hot water network. The hot water feeder forms an automatic pressure safety barrier in heat supply of heating or technological steam from a nuclear installation. (Z.M.). 5 figs., 9 refs

  6. Reliability of CANDU heat transport pumps

    International Nuclear Information System (INIS)

    Earl, A.H.

    1984-01-01

    While CANDU heat transport pumps are effectively demonstrating their reliability in the most reliable reactors in the world, the current level of reliability was only achieved after ongoing refinements to the original design. Each design modification was initiated in response to a problem, with reliability taking precedence over cost and efficiency so that each modification has predictably increased pump reliability. Seal reliability, a problem with most large, high temperature pumps, has through an extensive research and development program been improved to a point where a reliable lifetime of about five years can be expected. Co-operative research and development programs at bearing manufacturers facilities and at AECL's Engineering Laboratory have resulted in no bearing related pump incapabilities at either Pickering or Bruce nuclear generating stations. Motor windings, a source of unreliability at the Bruce station, have been improved through the introduction of better turn insulation and additional surge protection. Refinements to the current pump design will continue and AECL is placing particular emphasis on evaluating pump performance in post-LOCA conditions. (orig.)

  7. Upgrading primary heat transport pump seals

    International Nuclear Information System (INIS)

    Graham, T.; Metcalfe, R.; Rhodes, D.; McInnes, D.

    1995-01-01

    Changes in the operating environment at the Bruce-A Nuclear Generating Station created the need for an upgraded Primary Heat Transport Pump (PHTP) seal. In particular, the requirement for low pressure running during more frequent start-ups exposed a weakness of the CAN2 seal and reduced its reliability. The primary concern at Bruce-A was the rotation of the CAN2 No. 2 stators in their holders. The introduction of low pressure running exacerbated this problem, giving rapid wear of the stator back face, overheating, and thermocracking. In addition, the resulting increase in friction between the stator and its holder increased stationary-side hysteresis and thereby changed the seal characteristic to the point where interseal pressure oscillations became prevalent. The resultant increased hysteresis also led to hard rubbing of the seal faces during temperature transients. An upgraded seal was required for improved reliability to avoid forced outages and to reduce maintenance costs. This paper describes this upgraded 'replacement seal' and its performance history. In spite of the 'teething' problems detailed in this paper, there have been no forced outages due to the replacement seal, and in the words of a seal maintenance worker at Bruce-A, 'it allows me to go home and sleep at night instead of worrying about seal failures.' (author)

  8. Ductile fracture behaviour of primary heat transport piping material ...

    Indian Academy of Sciences (India)

    R. Narasimhan (Krishtel eMaging) 1461 1996 Oct 15 13:05:22

    Heat transport piping material of nuclear reactors. 169. Table 1. Chemical composition of. SA333, Grade 6 steel in wt%. C. Mn. Si. S. P. 0·18 0·90 0·25 0·02 0·02 room temperature. The tensile flow curve of the steel exhibited prominent yield-point effect accompanied by non-hardening strain propagation (Lüders strain) of ...

  9. Tritium transport around nuclear facilities

    International Nuclear Information System (INIS)

    Murphy, C.E. Jr.; Sweet, C.W.

    1981-01-01

    The transport and cycling of tritium around nuclear facilities is reviewed with special emphasis on studies at the Savannah River Laboratory, Aiken, South Carolina. These studies have shown that the rate of deposition from the atmosphere, the site of deposition, and the subsequent cycling are strongly influenced by the compound with which the tritium is associated. Tritiated hydrogen is largely deposited in the soil, while tritiated water is deposited in the greatest quantity in the vegetation. Tritiated hydrogen is converted in the soil to tritiated water that leaves the soil slowly, through drainage and transpiration. Tritiated water deposited directly to the vegetation leaves the vegetation more rapidly after exposure. Only a small part of the tritium entering the vegetation becomes bound in organic molecules. However, it appears tht the existence of soil organic compounds with tritium concentrations greater than the equilibrium concentration in the associated water can be explained by direct metabolism of tritiated hydrogen in vegetation

  10. The prospects for nuclear heating in Hungary

    International Nuclear Information System (INIS)

    Lynch, G.F.; Papp, I.

    1989-09-01

    Hungary supplies only half of its energy requirements from domestic resources and is very dependent upon imports of oil, natural gas and electricity to meet the current demand. In planning to reduce the dependence on imports, nuclear technology is considered an important element in the long-term energy strategy. To this end, an aggressive nuclear electricity generation program is being implemented with four 440 MWe units now operating and two 1000 MWe units committed. However, nuclear technology must be used in other energy sectors if the goal of long-term energy independence is to be achieved. On the demand side, 30% of the primary energy is consumed in the public sector, the major component being residential heating. Of the 3.7 million apartments in Hungary, 500 000 benefit from being connected to municipal district heating systems that use natural gas or oil as the energy base. This is, therefore, another significant energy sector that is amenable to using nuclear technology to substitute for imported oil and natural gas. In assessing alternative nuclear heat sources, a joint study was undertaken between Canada and Hungary to determine the feasibility of using the SLOWPOKE Energy System that has recently been developed. The SLOWPOKE Energy System is a benign nuclear heat source designed to supply 10 thermal megawatts in the form of hot water for local heating systems in buildings and institutions. It uses a combination of inherent safety features, including natural convection circulation and negative reactivity coefficients, and engineered features to ensure an extremely safe system. A SLOWPOKE demonstration heating reactor has been constructed in Canada. The unit started operation in 1987 July and is currently undergoing an extensive test program

  11. Advances in Nuclear Power Process Heat Applications

    International Nuclear Information System (INIS)

    2012-05-01

    Following an IAEA coordinated research project, this publication compiles the findings of research and development activities related to practical nuclear process heat applications. An overview of current progress on high temperature gas cooled reactors coupling schemes for different process heat applications, such as hydrogen production and desalination is included. The associated safety aspects are also highlighted. The summary report documents the results and conclusions of the project.

  12. High temperature nuclear heat for isothermal reformer

    International Nuclear Information System (INIS)

    Epstein, M.

    2000-01-01

    High temperature nuclear heat can be used to operate a reformer with various feedstock materials. The product synthesis gas can be used not only as a source for hydrogen and as a feedstock for many essential chemical industries, such as ammonia and other products, but also for methanol and synthetic fuels. It can also be burnt directly in a combustion chamber of a gas turbine in an efficient combined cycle and generate electricity. In addition, it can be used as fuel for fuel cells. The reforming reaction is endothermic and the contribution of the nuclear energy to the calorific value of the final product (synthesis gas) is about 25%, compared to the calorific value of the feedstock reactants. If the feedstock is from fossil origin, the nuclear energy contributes to a substantial reduction in CO 2 emission to the atmosphere. The catalytic steam reforming of natural gas is the most common process. However, other feedstock materials, such as biogas, landfill gas and CO 2 -contaminated natural gas, can be reformed as well, either directly or with the addition of steam. The industrial steam reformers are generally fixed bed reactors, and their performance is strongly affected by the heat transfer from the furnace to the catalyst tubes. In top-fired as well as side-fired industrial configurations of steam reformers, the radiation is the main mechanism of heat transfer and convection heat transfer is negligible. The flames and the furnace gas constitute the main sources of the heat. In the nuclear reformers developed primarily in Germany, in connection with the EVA-ADAM project (closed cycle), the nuclear heat is transferred from the nuclear reactor coolant gas by convection, using a heating jacket around the reformer tubes. In this presentation it is proposed that the helium in a secondary loop, used to cool the nuclear reactor, will be employed to evaporate intermediate medium, such as sodium, zinc and aluminum chloride. Then, the vapors of the medium material transfer

  13. Economic feasibility of heat supply from nuclear power plants in the United States

    International Nuclear Information System (INIS)

    Roe, K.K.; Oliker, I.

    1987-01-01

    Nuclear energy is regarded as competitive for urban district heating applications. Hot water heat transoport systems of up to 50 miles are feasible for heat loads over 1500 MWt, and heat load density of over 130 MWt/mi 2 is most suitable for nuclear applications. An incremental approach and a nuclear plant design provision for future heat extraction are recommended. Nuclear district heating technology status is discussed, particularly turbine design. Results of a study for retrofitting a major existing nuclear power plant to cogeneration operation are presented. The study indicates that for transmission distances up to 20 miles it is economical to generate and transport between 600 and 1200 MWt of district heat (author)

  14. Nuclear heating of oil-bearing formations

    International Nuclear Information System (INIS)

    Alspaw, D.I.

    1981-01-01

    The invention reported consists of reprocessing spent nuclear fuel to obtain fission products which are packaged and emplaced in salt water underlying viscous oil formations. The heat generated is transferred to the oil formation by convection and the viscosity of the oil is decreased enabling recovery from a productioon well. Improvement over prior art consists in emplacing the fission-product packages in the water formation, hence preventing charring of the oil near the heat source. If charring does occur, recovery of petroleum becomes even more difficult than prior to the application of heat. (O.T.)

  15. 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)

  16. Management of the process of nuclear transport

    International Nuclear Information System (INIS)

    Requejo, P.

    2015-01-01

    Since 1996 ETSA is the only Spanish logistics operator specialized on servicing the nuclear and radioactive industry. Nowadays ETSA has some technological systems specifically designed for the management of nuclear transports. These tools have been the result of the analysis of multiple factors involved in nuclear shipments, of ETSAs wide experience as a logistics operator and the search for continuous improvement. (Author)

  17. Nuclear fuel transport and particularly spent fuel transport

    International Nuclear Information System (INIS)

    Lenail, B.

    1986-01-01

    Nuclear material transport is an essential activity for COGEMA linking the different steps of the fuel cycle transport systems have to be safe and reliable. Spent fuel transport is more particularly examined in this paper because the development of reprocessing plant. Industrial, techmical and economical aspects are reviewed [fr

  18. ECRH and electron heat transport in tokamaks

    International Nuclear Information System (INIS)

    Zou, X.L.; Giruzzi, G.; Dumont, R.J.

    2003-01-01

    It has been observed during the ECRH experiments in tokamaks that the shape of the electron temperature profile in stationary regimes is not very sensitive to the ECRH power deposition i.e. the temperature profile remains peaked at the center even though the ECRH power deposition is off-axis. Various models have been invoked for the interpretation of this profile resilience phenomenon: the inward heat pinch, the critical temperature gradient, the Self-Organized Criticality, etc. Except the pinch effect, all of these models need a specific form of the diffusivity in the heat transport equation. In this work, our approach is to solve a simplified time-dependent heat transport equation analytically in cylindrical geometry. The features of this analytical solution are analyzed, in particular the relationship between the temperature profile resilience and the Eigenmode of the physical system with respect to the heat transport phenomenon. Finally, applications of this analytical solution for the determination of the transport coefficient and the polarization of the EC waves are presented. It has been shown that the solution of the simplified transport equation in a finite cylinder is a Fourier-Bessel series. This series represents in fact a decomposition of the heat source in Eigenmode, which are characterized by the Bessel functions of order 0. The physical interpretation of the Eigenmodes is the following: when the heat source is given by a Bessel function of order 0, the temperature profile has exactly the same form as the source at every time. At the beginning of the power injection, the effectiveness of the temperature response is the same for each Eigenmode, and the response in temperature, having the same form as the source, is local. Conversely, in the later phase of the evolution, the effectiveness of the temperature response for each Eigenmode is different: the higher the order, the lower the effectiveness. In this case the response in temperature appears as

  19. Desalination of sea water using nuclear heat

    International Nuclear Information System (INIS)

    Raisic, N.

    1977-01-01

    At present, desalination is the only major unconventional source of water supply of economic significance. It is the only alternative that is technically feasible with respect to large-scale operation. The total world desalination capacity is about 2.1 million cubic metres per day, which is indeed a marginal addition to conventional water supplies. There are two approaches to the use of nuclear reactors in desalination: they can be used for a single purpose (i.e. heat production) or for dual purpose (generating electricity and heat production). For either application, the nuclear reactor must meet several requirements: (1) it should be available in sizes of practical interest for dual power/ desalination or for desalination only; (2) the technology of the particular reactor type should be well developed to guarantee safe and reliable operation; (3) the economics of heat production should be attractive and competitive with other available energy sources

  20. Transport device for nuclear fuel powder

    International Nuclear Information System (INIS)

    Adelmann, M.

    1987-01-01

    The transport device for nuclear fuel powder, which does not disintegrate during transport, has a transport pipe which starts with its entry end from the floor or a closed container and opens with its outlet end at the top into a closed separation container connect via a powder filter to a suction pump. By alternate regular opening and closing of a first control valve for transport gas fitted to a transport pipe to a supply duct and a second control valve for transport gas fitted to the container to an additional supply duct, alternating plugs of nuclear fuel powder and transport gas cushions are formed and are transported to the outlet end of the transport pipe. (orig./HP) [de

  1. 75 MW heat extraction from Beznau nuclear power plant (Switzerland)

    International Nuclear Information System (INIS)

    Handl, K.H.

    1998-01-01

    The district heat extraction system installed and commissioned at the Beznau Nuclear Power Plant 1983 and 1984 is working successfully since the beginning. Together with a six kilometres extension in 1994, the system now consists of a 35 kilometres main network and 85 kilometres of local distribution pipelines. The eight founding communities as well as three networks joined later have been connected. Today around 2160 consumers of the Refuna district heating, small and large private buildings, industrial and agricultural enterprises are supplied with heat from the Beznau plant (1997: 141'000 MWh). The regional district heat supply system has become an integrated part of the regional infrastructure for around 20'000 inhabitants of the lower Aare valley. Nearly 15 years of operational experience are confirming the success of the strict approval conditions for the housing connections. Remarkably deep return flow temperatures in the district heating network were leading to considerable reserves in the transport capacity of the main pipeline system. The impacts of the heat extraction from the Beznau nuclear power plant, in particular its contribution to the protection of the environment by substituting fossil fuels and preventing CO2-production, have been positive. (author)

  2. Heat Transfer in Directional Water Transport Fabrics

    Directory of Open Access Journals (Sweden)

    Chao Zeng

    2016-10-01

    Full Text Available Directional water transport fabrics can proactively transfer moisture from the body. They show great potential in making sportswear and summer clothing. While moisture transfer has been previously reported, heat transfer in directional water transport fabrics has been little reported in research literature. In this study, a directional water transport fabric was prepared using an electrospraying technique and its heat transfer properties under dry and wet states were evaluated, and compared with untreated control fabric and the one pre-treated with NaOH. All the fabric samples showed similar heat transfer features in the dry state, and the equilibrium temperature in the dry state was higher than for the wet state. Wetting considerably enhanced the thermal conductivity of the fabrics. Our studies indicate that directional water transport treatment assists in moving water toward one side of the fabric, but has little effect on thermal transfer performance. This study may be useful for development of “smart” textiles for various applications.

  3. Miniature Heat Transport System for Nanosatellite Technology

    Science.gov (United States)

    Douglas, Donya M,

    1999-01-01

    The scientific understanding of key physical processes between the Sun and the Earth require simultaneous measurements from many vantage points in space. Nano-satellite technologies will enable a class of constellation missions for the NASA Space Science Sun-Earth Connections. This recent emphasis on the implementation of smaller satellites leads to a requirement for development of smaller subsystems in several areas. Key technologies under development include: advanced miniaturized chemical propulsion; miniaturized sensors; highly integrated, compact electronics; autonomous onboard and ground operations; miniatures low power tracking techniques for orbit determination; onboard RF communications capable of transmitting data to the ground from far distances; lightweight efficient solar array panels; lightweight, high output battery cells; lightweight yet strong composite materials for the nano-spacecraft and deployer-ship structures. These newer smaller systems may have higher power densities and higher thermal transport requirements than seen on previous small satellites. Furthermore, the small satellites may also have a requirement to maintain thermal control through extended earth shadows, possibly up to 8 hours long. Older thermal control technology, such as heaters, thermostats, and heat pipes, may not be sufficient to meet the requirements of these new systems. Conversely, a miniature two-phase heat transport system (Mini-HTS) such as a Capillary Pumped Loop (CPL) or Loop Heat Pipe (LBP) is a viable alternative. A Mini-HTS can provide fine temperature control, thermal diode action, and a highly efficient means of heat transfer. The Mini-HTS would have power capabilities in the range of tens of watts or less and provide thermal control over typical spacecraft ranges. The Mini-HTS would allow the internal portion of the spacecraft to be thermally isolated from the external radiator, thus protecting the internal components from extreme cold temperatures during an

  4. A five MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Dafang; Don Duo; Su Quingshan

    1997-01-01

    The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy and Technology (INET) and has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection, and environmental impacts and so on, were also obtained at the same time. All of these demonstrate that the design of NHR-5 is successful. (author)

  5. Hydrogen and oxygen production with nuclear heat

    International Nuclear Information System (INIS)

    Barnert, H.

    1979-09-01

    After some remarks on the necessity of producing secondary energy sources for the heat market, the thermodynamic fundamentals of the processes for producing hydrogen and oxygen from water on the basis of nuclear thermal energy are briefly explained. These processes are summarized as one class of the 'thermochemical cycle process' for the conversion of thermal into chemical energy. A number of thermochemical cycle processes are described. The results of the design work so far are illustrated by the example of the 'sulphuric acid hybrid process'. The nuclear heat source of the thermochemical cycle process is the high-temperature reactor. Statements concerning rentability are briefly commented upon, and the research and development efforts and expenditure required are sketched. (orig.) 891 GG/orig. 892 MB [de

  6. Transport of encapsulated nuclear fuels

    International Nuclear Information System (INIS)

    Broman, Ulrika; Dybeck, Peter; Ekendahl, Ann-Mari

    2005-12-01

    The transport system for encapsulated fuel is described, including a preliminary drawing of a transport container. In the report, the encapsulation plant is assumed to be located to Oskarshamn, and the repository to Oskarshamn or Forsmark

  7. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

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

  8. Transport packages for nuclear material and waste

    International Nuclear Information System (INIS)

    1997-01-01

    The regulations and responsibilities concerning the transport packages of nuclear materials and waste are given in the guide. The approval procedure, control of manufacturing, commissioning of the packaging and the control of use are specified. (13 refs.)

  9. One-Loop Operation of Primary Heat Transport System in MONJU During Heat Transport System Modifications

    International Nuclear Information System (INIS)

    Goto, T.; Tsushima, H.; Sakurai, N.; Jo, T.

    2006-01-01

    MONJU is a prototype fast breeder reactor (FBR). Modification work commenced in March 2005. Since June 2004, MONJU has changed to one-loop operation of the primary heat transport system (PHTS) with all of the secondary heat transport systems (SHTS) drained of sodium. The purposes of this change are to shorten the modification period and to reduce the cost incurred for circuit trace heating electrical consumption. Before changing condition, the following issues were investigated to show that this mode of operation was possible. The heat loss from the reactor vessel and the single primary loop must exceed the decay heat by an acceptable margin but the capacity of pre-heaters to keep the sodium within the primary vessel at about 200 deg. C must be maintained. With regard to the heat loss and the decay heat, the estimated heat loss in the primary system was in the range of 90-170 kW in one-loop operation, and the calculated decay heat was 21.2 kW. Although the heat input of the primary pump was considered, it was clear that circuit heat loss greatly exceeded the decay heat. As for pre-heaters, effective capacity was less than the heat loss. Therefore, the temperature of the reactor vessel room was raised to reduce the heat loss. One-loop operation of the PHTS was able to be executed by means of these measures. The cost of electrical consumption in the power plant has been reduced by one-loop operation of the PHTS and the modification period was shortened. (authors)

  10. Transport and reprocessing of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Lenail, B.

    1981-01-01

    This contribution deals with transport and packaging of oxide fuel from and to the Cogema reprocessing plant at La Hague (France). After a general discussion of nuclear fuel and the fuel cycle, the main aspects of transport and reprocessing of oxide fuel are analysed. (Auth.)

  11. Qualification of γ-heating calculation in nuclear reactors

    International Nuclear Information System (INIS)

    Ravaux, Simon

    2013-01-01

    During the last few years, the γ-heating issue has gained in stature, mainly for the safety of the 3. generation reactors in which a stainless steel reflector is inserted. The purpose of this work is the qualification of the needed tools for calculation of the γ-heating in the nuclear reactors. In a nuclear reactor, all the photons are directly or indirectly produced by the neutron-matter interactions. Thus, the first phase of this work is a critical analysis of the photon production data in the standard nuclear data library. New evaluations have been proposed to the next version of the JEFF library after that some omissions have been found. They have partly been accepted for JEFF-3.2. Two particle-transport codes are currently developed in the CEA: the deterministic code APOLLO2 and the Monte Carlo code TRIPOLI4. The second part of this work is the qualification of both these codes by interpreting an integral experiment called PERLE. The experimental set-up is made by a LWR pin assembly surrounded by a stainless steel reflector in which the γ-heating is measured by Thermo-luminescent Detector (TLD). A calculation scheme has been proposed for both APOLLO2 and TRIPOLI4 in order to calculate the TLD's responses. Comparisons between calculations and measurements have shown that TRIPOLI4 gives a satisfactory estimation of the γ-heating in the reflector. These discrepancies are within the experimental 1 σ uncertainty. Before the qualification, APOLLO2 has been previously validated against TRIPOLI4 reference calculation. This validation gives an estimation of the bias due to the deterministic approximations of the transport equation resolution. The qualification has shown that the discrepancies between APOLLO2 predictions and TLD's measurements are in the same range as experimental uncertainties. (author) [fr

  12. The Next Nuclear Gamble. Transportation and storage of nuclear waste

    International Nuclear Information System (INIS)

    Resnikoff, M.

    1985-01-01

    The Next Nuclear Gamble examines risks, costs, and alternatives in handling irradiated nuclear fuel. The debate over nuclear power and the disposal of its high-level radioactive waste is now nearly four decades old. Ever larger quantities of commercial radioactive fuel continue to accumulate in reactor storage pools throughout the country and no permanent storage solution has yet been designated. As an interim solution, the government and utilities prefer that radioactive wastes be transported to temporary storage facilities and subsequently to a permanent depository. If this temporary and centralized storage system is implemented, however, the number of nuclear waste shipments on the highway will increase one hundredfold over the next fifteen years. The question directly addressed is whether nuclear transport is safe or represents the American public's domestic nuclear gamble. This Council on Economic Priorities study, directed by Marvin Resnikoff, shows on the basis of hundreds of government and industry reports, interviews and surveys, and original research, that transportation of nuclear materials as currently practiced is unsafe

  13. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  14. Legal aspects of transport of nuclear materials

    International Nuclear Information System (INIS)

    Jacobsson, Mans.

    The Paris Convention and the Brussels Supplementary Convention are briefly discussed and other conventions in the field of civil liability for nuclear damage are mentioned: the Vienna Convention, the Nuclear Ships Convention and the 1971 Convention relating to civil liability in the field of maritime carriage of nuclear material. Legislation on civil liability in the Nordic countries, which is based on the Paris Convention and the Supplementary Convention is discussed, notably the principle of channelling of liability and exceptions from that principle due to rules of liability in older transport conventions and certain problems due to the limited geographical scope of the Paris Convention and the Supplementary Convention. Insurance problems arising in connection with transport of nuclear materials are surveyed and an outline is given of the administrative provisions concerning transport (based on the IAEA transport regulations) which govern transport of radioactive materials by different means: road, rail, sea and air. Finally, the 1968 Treaty on the Non-Proliferation of Nuclear Weapons is discussed. (NEA) [fr

  15. Offshore heat dissipation for nuclear energy centers

    International Nuclear Information System (INIS)

    Bauman, H.F.

    1978-09-01

    The technical, environmental, and economic aspects of utilizing the ocean or other large water bodies for the dissipation of reject heat from Nuclear Energy Centers (NECs) were investigated. An NEC in concept is an aggregate of nuclear power plants of 10 GW(e) capacity or greater on a common site. The use of once-through cooling for large power installations offers advantages including higher thermal efficiencies, especially under summer peak-load conditions, compared to closed-cycle cooling systems. A disadvantage of once-through cooling is the potential for greater adverse impacts on the aquatic environment. A concept is presented for minimizing the impacts of such systems by placing water intake and discharge locations relatively distant from shore in deeper water than has heretofore been the practice. This technique would avoid impacts on relatively biologically productive and ecologically sensitive shallow inshore areas. The NEC itself would be set back from the shoreline so that recreational use of the shore area would not be impaired. The characteristics of a heat-dissipation system of the size required for a NEC were predicted from the known characteristics of a smaller system by applying hydraulic scaling laws. The results showed that adequate heat dissipation can be obtained from NEC-sized systems located in water of appropriate depth. Offshore intake and discharge structures would be connected to the NEC pump house on shore via tunnels or buried pipelines. Tunnels have the advantage that shoreline and beach areas would not be disturbed. The cost of an offshore heat-dissipation system depends on the characteristics of the site, particularly the distance to suitably deep water and the type of soil or rock in which water conduits would be constructed. For a favorable site, the cost of an offshore system is estimated to be less than the cost of a closed-cycle system

  16. Current state of utilization of nuclear sources for district heating

    International Nuclear Information System (INIS)

    Zlatnansky, J.

    1987-01-01

    The use of heat from nuclear power sources is considered as the basic trend of development in district heating within the framework of the long-term prospective plan of the development of the fuel and power complex in the CSSR. Of the total consumption of heat in the year 2005 (648,000 TJ), nuclear power units are to deliver 54,000 TJ. At present, heat is being generated by the Jaslovske Bohunice nuclear power plant and supplied to Trnava, and a heat duct is being built from this plant to Leopoldov and Hlohovce, and prospectively to Bratislava. The Dukovany nuclear power plant is expected to supply heat to Brno, the Mochovce nuclear power plant to Levice and Nitra, and the Temelin nuclear power plant to Ceske Budejovice. Other nuclear power plants will in the future also supply heat (Kecerovce, Blahutovice and Opatovice). The technical and economic evaluation is tabulated of heat deliveries from all said nuclear power plants. It is stated that after the year 2000, AST 500 nuclear heating plants should be deployed for supplying heat to areas out of reach of nuclear power plants and reconstructed fossil-fuel power plants. (Z.M.). 2 tabs

  17. Nuclear heat applications in Russia: Experience, status and prospects

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kusmartsev, E.V.

    1998-01-01

    The extensive experience gained with nuclear district heating in Russia is described. Most of the WWER reactors in Russia are cogeneration plants. Steam is extracted through LP turbine bleeders and condensed in intermediate heat exchangers to hot water which is then supplied to DH grids. Also some small dedicated nuclear heating plants are operated. (author)

  18. Transporting spent nuclear fuel: an overview

    International Nuclear Information System (INIS)

    1986-03-01

    Although high-level radioactive waste from both commercial and defense activities will be shipped to the repository, this booklet focuses on various aspects of transporting commercial spent fuel, which accounts for the majority of the material to be shipped. The booklet is intended to give the reader a basic understanding of the following: the reasons for transportation of spent nuclear fuel, the methods by which it is shipped, the safety and security precautions taken for its transportation, emergency response procedures in the event of an accident, and the DOE program to develop a system uniquely appropriate to NWPA transportation requirements

  19. Next nuclear gamble: transportation and storage of nuclear waste

    International Nuclear Information System (INIS)

    Resnikoff, M.

    1983-01-01

    Accidents during transport of nuclear waste are more threatening - though less likely - than a reactor meltdown because transportation accidents could occur in the middle of a populous city, affecting more people and property than a plant accident, according to the Council on Economic Priorities, a non-profit public service research organization. Transportation, as presently practiced, is unsafe. Shipping containers, called casks, are poorly designed and constructed, CEP says. The problem needs attention because the number of casks filled with nuclear waste on the nation's highways could increase a hundred times during the next 15 years under the Nuclear Waste Policy Act of 1982, which calls for storage areas. Recommendations, both technical and regulatory, for reducing the risks are presented

  20. Management of the process of nuclear transport; Gestion del proceso de transporte nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Requejo, P.

    2015-07-01

    Since 1996 ETSA is the only Spanish logistics operator specialized on servicing the nuclear and radioactive industry. Nowadays ETSA has some technological systems specifically designed for the management of nuclear transports. These tools have been the result of the analysis of multiple factors involved in nuclear shipments, of ETSAs wide experience as a logistics operator and the search for continuous improvement. (Author)

  1. Overview of nuclear materials transportation

    International Nuclear Information System (INIS)

    Grella, A.W.

    1986-01-01

    This presentation is an overview of transportation as it relates to one specific type of material, low specific activity (LSA) material. It is the predominant type of material that fits into the low-level waste category. An attempt is made to discuss how LSA is regulated, setting forth the requirements. First the general scheme of regulations are reviewed. In addition future changes in the regulations which will affect transportation of LSA materials and, which quite likely, will have an impact on R and D needs in this area are presented

  2. GLONASS satellite monitoring of nuclear transports

    International Nuclear Information System (INIS)

    Davydov, Yu.L.

    2012-01-01

    In 2011 Rosatom has made the decision to create the industry-wide automated system for monitoring of transports of radioactive substances (RS) and wastes (RAW), as well as hazardous loads by rail and automobile, based upon the same hardware as used by the GLONASS satellite navigation system - the so-called ASBT-GLONASS system. The new system will use the same technical infrastructure as the existing operational Automated System for Safe Transport of Nuclear Materials of Categories I and II (ASBT). The ASBT structure includes a network of control centres fitted with automation and communication hardware. In addition, ASBT includes technical complexes installed upon transport vehicles intended for nuclear material transport. In order to identify transport vehicle location, the GLONASS/GPS (GALS-P-ASBT) satellite navigational receiver device is used, it is developed especially for ASBT systems taking in account information security requirements. By now the basic software and hardware complex ASBT-GLONASS has been created (equipment to be carried on-board the transport vehicle loaded with RS and RAW, as well as the transport control stations) that supports transport monitoring and transmission of an emergency signal to control stations of companies which deal with RS and RAW transportation [ru

  3. Design of SES-10 nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Cuttler, J.M.

    1991-03-01

    The SES-10 units are unpressurized, pool-type nuclear reactors of 10MW rating, designed for supplying energy to hot water district heating systems, economically and without pollution. Water for heat distribution is brought to a maximum temperature of 85 degrees C. Conventional heating units supplement the output from SES-10 units for peak load and during maintenance. The SES-10 is housed in a low-cost building, with a double-walled pool in the ground. A naturally circulating primary system and a pumped secondary system transport heat from the reactor to the distribution system. The unit is fully automated and easy to maintain. Because of the many active and passive safety features, it is feasible to license the SES-10 for operation in a city and easy to explain it to the public for their acceptance. The core lasts approximately 43 months at a capacity factor of 70%, and the cost of heat is expected to be 2 to 2.5 cents/kWh

  4. Design of SES-10 nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Cuttler, J.M.

    1991-01-01

    The SES-10 units are unpressurized, pool-type nuclear reactors of 10 MW rating, designed for supplying energy to hot water district heating systems, economically and without pollution. Water for heat distribution is brought to a maximum temperature of 85 o C. Conventional heating units supplement the output from SES-10 units for peak load and during maintenance. The SES-10 is housed in a low-cost building, with a double-walled pool in the ground. A naturally circulating primary system and a pumped secondary system transport heat from the reactor to the distribution system. The unit is fully automated and easy to maintain. Because of the many active and passive safety features, it is feasible to license the SES-10 for operation in a city and easy to explain it to the public for their acceptance. The core lasts approximately 43 months at a capacity factor of 70%, and the cost of heat is expected to be 2 to 2.5 cents/kWh. (author) 8 figs

  5. Uncontrolled transport of nuclear materials

    International Nuclear Information System (INIS)

    Wassermann, U.

    1985-01-01

    An account is given of international transport of plutonium, uranium oxides, uranium hexafluoride, enriched uranium and irradiated fuel for reprocessing. Referring to the sinking of the 'Mont Louis', it is stated that the International Maritime Organization has been asked by the National Union of Seamen and 'Greenpeace' to bar shipment of radioactive material until stricter international safety regulations are introduced. (U.K.)

  6. Nuclear power plant waste heat utilization

    International Nuclear Information System (INIS)

    Ryther, J.H.; Huke, R.E.; Archer, J.C.; Price, D.R.; Jewell, W.J.; Hayes, T.D.; Witherby, H.R.

    1977-09-01

    The possibility of using Vermont Yankee condenser effluent for commercial food growth enhancement was examined. It was concluded that for the Vermont Yankee Nuclear Station, commercial success, both for horticulture and aquaculture endeavors, could not be assured without additional research in both areas. This is due primarily to two problems. First, the particularly low heat quality of our condenser discharge, being nominally 72 +- 2 0 F; and second, to the capital intensive support systems. The capital needed for the support systems include costs of pumps, piping and controls to move the heated water to growing facilities and the costs of large, efficient heat exchangers that may be necessary to avoid regulatory difficulties due to the 1958 Delaney Amendment to the U.S. Food, Drug and Cosmetics Act. Recommendations for further work include construction of a permanent aquaculture research laboratory and a test greenhouse complex based on a greenhouse wherein a variety of heating configurations would be installed and tested. One greenhouse would be heated with biogas from an adjacent anaerobic digester thermally boosted during winter months by Vermont Yankee condenser effluent. The aquaculture laboratory would initially be dedicated to the Atlantic salmon restoration program. It appears possible to raise fingerling salmon to smolt size within 7 months using water warmed to about 60 0 F. The growth rate by this technique is increased by a factor of 2 to 3. A system concept has been developed which includes an aqua-laboratory, producing 25,000 salmon smolt annually, a 4-unit greenhouse test horticulture complex and an 18,000 square foot commercial fish-rearing facility producing 100,000 pounds of wet fish (brook trout) per year. The aqualab and horticulture test complex would form the initial phase of construction. The trout-rearing facility would be delayed pending results of laboratory studies confirming its commercial viability

  7. Nuclear power plant waste heat utilization

    Energy Technology Data Exchange (ETDEWEB)

    Ryther, J.H.; Huke, R.E.; Archer, J.C.; Price, D.R.; Jewell, W.J.; Hayes, T.D.; Witherby, H.R.

    1977-09-01

    The possibility of using Vermont Yankee condenser effluent for commercial food growth enhancement was examined. It was concluded that for the Vermont Yankee Nuclear Station, commercial success, both for horticulture and aquaculture endeavors, could not be assured without additional research in both areas. This is due primarily to two problems. First, the particularly low heat quality of our condenser discharge, being nominally 72 +- 2/sup 0/F; and second, to the capital intensive support systems. The capital needed for the support systems include costs of pumps, piping and controls to move the heated water to growing facilities and the costs of large, efficient heat exchangers that may be necessary to avoid regulatory difficulties due to the 1958 Delaney Amendment to the U.S. Food, Drug and Cosmetics Act. Recommendations for further work include construction of a permanent aquaculture research laboratory and a test greenhouse complex based on a greenhouse wherein a variety of heating configurations would be installed and tested. One greenhouse would be heated with biogas from an adjacent anaerobic digester thermally boosted during winter months by Vermont Yankee condenser effluent. The aquaculture laboratory would initially be dedicated to the Atlantic salmon restoration program. It appears possible to raise fingerling salmon to smolt size within 7 months using water warmed to about 60/sup 0/F. The growth rate by this technique is increased by a factor of 2 to 3. A system concept has been developed which includes an aqua-laboratory, producing 25,000 salmon smolt annually, a 4-unit greenhouse test horticulture complex and an 18,000 square foot commercial fish-rearing facility producing 100,000 pounds of wet fish (brook trout) per year. The aqualab and horticulture test complex would form the initial phase of construction. The trout-rearing facility would be delayed pending results of laboratory studies confirming its commercial viability.

  8. Nuclear power generation and global heating

    International Nuclear Information System (INIS)

    Taboada, Horacio

    1999-01-01

    The Professionals Association and Nuclear Activity of National Atomic Energy Commission (CNEA) are following with great interest the worldwide discussions on global heating and the role that nuclear power is going to play. The Association has an active presence, as part of the WONUC (recognized by the United Nations as a Non-Governmental Organization) in the COP4, which was held in Buenos Aires in November 1998. The environmental problems are closely related to human development, the way of power production, the techniques for industrial production and exploitation fields. CO 2 is the most important gas with hothouse effects, responsible of progressive climatic changes, as floods, desertification, increase of average global temperature, thermal expansion in seas and even polar casks melting and ice falls. The consequences that global heating will have on the life and economy of human society cannot be sufficiently emphasized, great economical impact, destruction of ecosystems, loss of great coast areas and complete disappearance of islands owing to water level rise. The increase of power retained in the atmosphere generates more violent hurricanes and storms. In this work, the topics presented in the former AATN Meeting is analyzed in detail and different technological options and perspectives to mitigate CO 2 emission, as well as economical-financial aspects, are explored. (author)

  9. Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

    Science.gov (United States)

    Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

    2017-12-01

    This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

  10. Utilising heat from nuclear waste for space heating

    International Nuclear Information System (INIS)

    Deacon, D.

    1982-01-01

    A heating unit utilising the decay heat from irradiated material comprises a storage envelope for the material associated with a heat exchange system, means for producing a flow of air over the heat exchange system to extract heat from the material, an exhaust duct capable of discharging the heated air to the atmosphere, and means for selectively diverting at least some of the heated air to effect the required heating. With the flow of air over the heat exchange system taking place by a natural thermosyphon process the arrangement is self regulating and inherently reliable. (author)

  11. Institutional issues affecting transportation of nuclear materials

    International Nuclear Information System (INIS)

    Reese, R.T.; Luna, R.E.

    1980-01-01

    The institutional issues affecting transportation of nuclear materials in the United States represent significant barriers to meeting future needs in the transport of radioactive waste materials to their ultimate repository. While technological problems which must be overcome to perform such movements seem to be within the state-of-the-art, the timely resolution of these institutional issues seems less assured. However, the definition of these issues, as attempted in this paper, together with systematic analysis of cause and possible solutions are the essential elements of the Transportation Technology Center's Institutional Issues Program

  12. Heat in the Barents Sea: transport, storage, and surface fluxes

    Directory of Open Access Journals (Sweden)

    L. H. Smedsrud

    2010-02-01

    Full Text Available A column model is set up for the Barents Sea to explore sensitivity of surface fluxes and heat storage from varying ocean heat transport. Mean monthly ocean transport and atmospheric forcing are synthesised and force the simulations. Results show that by using updated ocean transports of heat and freshwater the vertical mean hydrographic seasonal cycle can be reproduced fairly well.

    Our results indicate that the ~70 TW of heat transported to the Barents Sea by ocean currents is lost in the southern Barents Sea as latent, sensible, and long wave radiation, each contributing 23–39 TW to the total heat loss. Solar radiation adds 26 TW in the south, as there is no significant ice production.

    The northern Barents Sea receives little ocean heat transport. This leads to a mixed layer at the freezing point during winter and significant ice production. There is little net surface heat loss annually in the north. The balance is achieved by a heat loss through long wave radiation all year, removing most of the summer solar heating.

    During the last decade the Barents Sea has experienced an atmospheric warming and an increased ocean heat transport. The Barents Sea responds to such large changes by adjusting temperature and heat loss. Decreasing the ocean heat transport below 50 TW starts a transition towards Arctic conditions. The heat loss in the Barents Sea depend on the effective area for cooling, and an increased heat transport leads to a spreading of warm water further north.

  13. Current status of research and development for nuclear heating reactor in China

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo

    1987-01-01

    At present the coal is the main source for district heating in China. It results in serious problems for transportation and pollution. Nuclear district heating reactor can substitute the coal and supply the clear and ecenomic heat energy for the cities. A feasibility studies for a district heating reactor with the power of 450 MW(t) in Harbin were carried out. With cooperation of heating boilers heat demand of 1.2 million pupulation can be satisfied. 600 x 10 3 tons coal per year can be saved. The temperature of the heat grid is 130/70 deg C. The main parameters of the 450 MW(t) and 5 MW(t) heating reactors are given. The technical design, safety aspects, economic analysis and the stability of test loop are also discussed. (Liu)

  14. Transport description of damped nuclear reactions

    International Nuclear Information System (INIS)

    Randrup, J.

    1983-04-01

    Part I is an elementary introduction to the general transport theory of nuclear dynamics. It can be read without any special knowledge of the field, although basic quantum mechanics is required for the formal derivation of the general expression for the transport coefficients. The results can also be used in a wider context than the present one. Part II gives the student an up-to-date orientation about recent progress in the understanding of the angular-momentum variables in damped reactions. The emphasis is here on the qualitative understanding of the physics rather than the, at times somewhat tedious, formal derivations

  15. Source effects on impurity and heat transport in a tokamak

    International Nuclear Information System (INIS)

    Bennett, R.B.

    1980-12-01

    A recently developed generalization of neoclassical theory is extended here to study heat flux contributions to impurity transport, as well as the heat fluxes themselves. The theory accounts for the first four source moments, with external drags, which has been studied previously with either fewer moments or restricted to a collisional plasma. Conditions are established for which a momentum source may be used to modify the particle and heat transport. In the course of this work, the particle and heat transport is evaluated for a two species plasma with arbitrary plasma geometry, beta, and collisionality

  16. Liability for international nuclear transport: an overview

    International Nuclear Information System (INIS)

    Brown, O.F.; Horbach, N.

    2000-01-01

    Many elements can bear on liability for nuclear damage during transport. For example, liability may depend upon a number of facts that may be categorized as follows: shipment, origin or destination of the shipment, deviation from the planed route, temporary storage incidental to carriage; content of shipment, type of nuclear material involved, whether its origin is civilian or defence-related; sites of accident, number and type of territories damaged (i.e. potential conventions involved), applicable territorial limits, exclusive economic zone, high seas, etc.; nature of damages, personal injury, property damage, damage to the means of carriage, indirect damage, preventive measures, environmental cleanup or retrieval at seas, res communis, transboundary damages etc.; victims involved, nationality and domiciles of victims; jurisdiction, flag (for ships) or national registration (for aircraft) of the transporting vessel, courts of one or more states may have (or assert) jurisdiction to hear claims, and may have to determine what law to apply to a particular accident; applicable law, the applicability laws and/or international nuclear liability conventions; the extent to which any applicable convention has been implemented or modified by domestic legislation, conflicts with the 1982 Law of the Sea Convention or other applicable international agreements, and finally, also written agreements between installation operators and carriers can define applicable law as well as responsibilities. Harmonizing nuclear liability protection and applying it to additional international shipments would be facilitated by more countries being in treaty relations with each other as soon as possible. Adherence to an international convention by more countries (including China, Russia, the United States, etc.) would promote the open flow of services and advanced technology, and better facilitate international transport. The conventions protect the public, harmonize legislation in the

  17. Nuclear district heating system with a high-temperature reactor

    International Nuclear Information System (INIS)

    Schroeder, G.; Barnert, H.; Wischnewski, R.

    1978-01-01

    From the demand viewpoint, the connection of an installed nuclear thermal capacity of 290 MJ/s for district heating purposes would be possible in the central Ruhr District by 1982-1983. The nuclear district heating system is made up of several subsystems, for instance, a smaller size high-temperature reactor [500 MW(thermal)] as a nuclear heat-and-power plant and an interconnected district heating system with a feed temperature of 453 K (180 0 C). The expenditure for additional investments, additional fuel costs, and costs for substitute power capacity are charged to the thermal energy generation costs of the nuclear heat-and-power plant. For the nuclear district heating system, the district heating costs to the consumer will vary over wide limits, depending on local conditions, between 7.8 and 12.2 $/GJ at the commissioning date in 1983, assuming that all subsystems have to be newly installed. These costs can be lower than district heating costs in a conventional district heating system with fossil-fired heating stations

  18. NEP heat pipe radiators. [Nuclear Electric Propulsion

    Science.gov (United States)

    Ernst, D. M.

    1979-01-01

    This paper covers improvements of heat pipe radiators for the thermionic NEP design. Liquid metal heat pipes are suitable as spacecraft radiator elements because of high thermal conductance, low mass and reliability, but the NEP thermionic system design was too large and difficult to fabricate. The current integral collector-radiator design consisting of several layers of thermionic converters, the annular-tangential collector heat pipe, the radiator heat pipe, and the transition zone designed to minimize the temperature difference between the collector heat pipe and radiator heat pipe are described. Finally, the design of micrometeoroid armor protection and the fabrication of the stainless steel annular heat pipe with a tangential arm are discussed, and it is concluded that the heat rejection system for the thermionic NEP system is well advanced, but the collector-radiator heat pipe transition and the 8 to 10 m radiator heat pipe with two bends require evaluation.

  19. Energy Conversion Advanced Heat Transport Loop and Power Cycle

    Energy Technology Data Exchange (ETDEWEB)

    Oh, C. H.

    2006-08-01

    The Department of Energy and the Idaho National Laboratory are developing a Next Generation Nuclear Plant (NGNP) to serve as a demonstration of state-of-the-art nuclear technology. The purpose of the demonstration is two fold 1) efficient low cost energy generation and 2) hydrogen production. Although a next generation plant could be developed as a single-purpose facility, early designs are expected to be dual-purpose. While hydrogen production and advanced energy cycles are still in its early stages of development, research towards coupling a high temperature reactor, electrical generation and hydrogen production is under way. Many aspects of the NGNP must be researched and developed in order to make recommendations on the final design of the plant. Parameters such as working conditions, cycle components, working fluids, and power conversion unit configurations must be understood. Three configurations of the power conversion unit were demonstrated in this study. A three-shaft design with 3 turbines and 4 compressors, a combined cycle with a Brayton top cycle and a Rankine bottoming cycle, and a reheated cycle with 3 stages of reheat were investigated. An intermediate heat transport loop for transporting process heat to a High Temperature Steam Electrolysis (HTSE) hydrogen production plant was used. Helium, CO2, and an 80% nitrogen, 20% helium mixture (by weight) were studied to determine the best working fluid in terms cycle efficiency and development cost. In each of these configurations the relative component size were estimated for the different working fluids. The relative size of the turbomachinery was measured by comparing the power input/output of the component. For heat exchangers the volume was computed and compared. Parametric studies away from the baseline values of the three-shaft and combined cycles were performed to determine the effect of varying conditions in the cycle. This gives some insight into the sensitivity of these cycles to various

  20. Carbon nanostructured surfaces for enhanced heat transport

    NARCIS (Netherlands)

    Taha, T.J.

    2015-01-01

    The advancement of high performance thermal systems has stimulated interest in methods to improve heat transfer rates. Considerable efforts have been made to increase heat transfer rates by implementing passive convective heat transfer enhancement methods that require no direct consumption of

  1. The changing nature of nuclear transport

    International Nuclear Information System (INIS)

    Brobst, W.A.

    1976-01-01

    The IAEA's efforts in transport safety have been proven through 25 years of nearly accident-free transport, with no evidence of death or injury from the nuclear characteristics of those shipments. Much testing has been done over the last five years to verify the technical bases for the IAEA Regulations. Rather than being complacent with the past, we should instead see ourselves at a turning point for the solution of problems coming up in the next few years. The number of shipments will increase drastically and this will result in changes of risk levels. A number of critical problems will be discussed: (1) lack of public acceptance of nuclear shipment safety; (2) transport safeguards; (3) incompleteness in the IAEA package damage tests; (4) need for innovative technology for spent fuel casks; (5) reduction of radiation dose to the public; (6) quality assurance; (7) engineering analyses versus scale-model and full-scale testing; and (8) transport controls. A recommendation is made to the IAEA to set up immediately a study group to define these problems, list alternatives and options, and recommend corrective actions. (author)

  2. Current trends in nuclear material transportation

    International Nuclear Information System (INIS)

    Ravenscroft, Norman; Oshinowo, Franchone

    1997-01-01

    The business of radioactive material transportation has evolved considerably in the past 40 years. Current practices reflect extensive international experience in handling radioactive cargo within a mature and tested regulatory framework. Nevertheless, new developments continue to have an impact on how shipments of nuclear material are planned and carried out. Entities involved in the transport of radioactive materials must keep abreast of these developments and work together to find innovative solutions to ensure that safe, smooth transport activities may continue. Several recent trends in the regulatory environment and political atmosphere require attention. There are four key trends that we'll be examining today: 1) the reduction in the pool of available commercial carriers; 2) routing restrictions; 3) package validation issues; and 4) increasing political sensitivities. Careful planning and cooperative measures are necessary to alleviate problems in each of these areas. (author)

  3. Survey of heat-pipe application under nuclear environment

    International Nuclear Information System (INIS)

    Tsuyuzaki, Noriyoshi; Saito, Takashi; Okamoto, Yoshizo; Hishida, Makoto; Negishi, Kanji.

    1986-11-01

    Heat pipes today are employed in a wide variety of special heat transfer applications including nuclear reactor. In this nuclear technology area in Japan, A headway speed of the heat pipe application technique is not so high because of safety confirmation and investigation under each developing step. Especially, the outline of space craft is a tendency to increase the size. Therefore, the power supply is also tendency to increase the outlet power and keep the long life. Under SP-100 project, the development of nuclear power supply system which power is 1400 - 1600 KW thermal and 100 KW electric power is steadily in progress. Many heat pipes are adopted for thermionic conversion and coolant system in order to construct more safety and light weight system for the project. This paper describes the survey of the heat pipe applications under the present and future condition for nuclear environment. (author)

  4. Allocation of fossil and nuclear fuels. Heat production from chemically and physically bound energy

    International Nuclear Information System (INIS)

    Wagner, U.

    2008-01-01

    The first part of the book presents the broad field of allocation, transformation, transport and distribution of the most important energy carriers in the modern power industry. The following chapters cover solid fossil fuel, liquid fuel, gaseous fuel and nuclear fuel. The final chapters concern the heat production from chemically and physically bound energy, including elementary analysis, combustion calculations, energy balance considerations in fossil fuel fired systems, and fundamentals of nuclear physics

  5. Economics of long distance transmission, storage and distribution of heat from nuclear plants with existing and newer techniques

    International Nuclear Information System (INIS)

    Margen, Peter

    1977-01-01

    Nuclear plants can provide heat for district heating in mainly two ways. Central nuclear power plants sufficiently large to be economic as electricity producers could instead be designed for heat extraction at temperatures useful for district heating. The second promising way is to design simple low temperature reactors, so simple and safe that near urban location becomes feasible. The manner of transport distribution and storage of heat is discussed in this paper which are very important especially in the cost calculations. The economic objectives can often be attained already with conventional technigues even when transport distances are large. But newer techniques of transport promise to make even cities at greater distances from major nuclear power plants economically connectible whilst new techniques for small distribution pipes help to extend the economic distribution area to the less dense one-family house districts. (M.S.)

  6. High temperature nuclear process heat systems for chemical processes

    International Nuclear Information System (INIS)

    Jiacoletti, R.J.

    1976-01-01

    The development planning and status of the very high temperature gas cooled reactor as a source of industrial process heat is presented. The dwindling domestic reserves of petroleum and natural gas dictate major increases in the utilization of coal and nuclear sources to meet the national energy demand. The nuclear process heat system offers a unique combination of the two that is environmentally and economically attractive and technically sound. Conceptual studies of several energy-intensive processes coupled to a nuclear heat source are presented

  7. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  8. Heat transfer and fluid flow in nuclear systems

    CERN Document Server

    Fenech, Henri

    1982-01-01

    Heat Transfer and Fluid in Flow Nuclear Systems discusses topics that bridge the gap between the fundamental principles and the designed practices. The book is comprised of six chapters that cover analysis of the predicting thermal-hydraulics performance of large nuclear reactors and associated heat-exchangers or steam generators of various nuclear systems. Chapter 1 tackles the general considerations on thermal design and performance requirements of nuclear reactor cores. The second chapter deals with pressurized subcooled light water systems, and the third chapter covers boiling water reacto

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

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

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

  10. Heat-pipe heat transport system for Stirling space power converter

    Science.gov (United States)

    Alger, Donald L.

    Life issues relating to a sodium-heat-pipe heat transport system are discussed. The heat-pipe system provides heat, at a temperature of 1050 K, to a 50-kWe Stirling engine/linear alternator power converter called the Stirling space power converter. Because corrosion of heat-pipe materials in contact with sodium can affect the life of the heat pipe, a literature review of sodium corrosion processes was performed. It was found that impurity reactions, primarily oxygen, and dissolution of alloy elements were the two corrosion processes likely to be operative in the heat pipe. Approaches that are being taken to minmize these corrosion processes are discussed.

  11. Experimental Study of Heat Transport in Fractured Network

    Science.gov (United States)

    Pastore, Nicola; Cherubini, Claudia; Giasi, Concetta I.; Allegretti, Nicoletta M.; Redondo, Jose M.; Tarquis, Ana Maria

    2015-04-01

    Fractured rocks play an important role in transport of natural resources or contaminants transport through subsurface systems. In recent years, interest has grown in investigating heat transport by means of tracer tests, driven by the important current development of geothermal applications. In literature different methods are available for predicting thermal breakthrough in fractured reservoirs based on the information coming from tracer tests. Geothermal energy is one of the largest sources of renewable energies that are extracted from the earth. The growing interest in this new energy source has stimulated attempts to develop methods and technologies for extracting energy also from ground resource at low temperature. An example is the exploitation of low enthalpy geothermal energy that can be obtained at any place with the aid of ground-source heat pump system from the soil, rock and groundwater. In such geothermal systems the fluid movement and thermal behavior in the fractured porous media is very important and critical. Existing theory of fluid flow and heat transport through porous media is of limited usefulness when applied to fractured rocks. Many field and laboratory tracer tests in fractured media show that fracture -matrix exchange is more significant for heat than mass tracers, thus thermal breakthrough curves (BTCs) are strongly controlled by matrix thermal diffusivity. In this study the behaviour of heat transport in a fractured network at bench scale has been investigated. Heat tracer tests on an artificially created fractured rock sample have been carried out. The observed thermal BTCs obtained with six thermocouple probes located at different locations in the fractured medium have been modeled with the Explicit Network Model (ENM) based an adaptation of Tang's solution for solute transport in a semi-infinite single fracture embedded in a porous matrix. The ENM model is able to represent the behavior of observed heat transport except where the

  12. Manned space flight nuclear system safety. Volume 4: Space shuttle nuclear system transportation. Part 1: Space shuttle nuclear safety

    Science.gov (United States)

    1972-01-01

    An analysis of the nuclear safety aspects (design and operational considerations) in the transport of nuclear payloads to and from earth orbit by the space shuttle is presented. Three representative nuclear payloads used in the study were: (1) the zirconium hydride reactor Brayton power module, (2) the large isotope Brayton power system and (3) small isotopic heat sources which can be a part of an upper stage or part of a logistics module. Reference data on the space shuttle and nuclear payloads are presented in an appendix. Safety oriented design and operational requirements were identified to integrate the nuclear payloads in the shuttle mission. Contingency situations were discussed and operations and design features were recommended to minimize the nuclear hazards. The study indicates the safety, design and operational advantages in the use of a nuclear payload transfer module. The transfer module can provide many of the safety related support functions (blast and fragmentation protection, environmental control, payload ejection) minimizing the direct impact on the shuttle.

  13. Titanium Loop Heat Pipes for Space Nuclear Radiators, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will develop titanium Loop Heat Pipes (LHPs) that can be used in low-mass space nuclear radiators, such as...

  14. Nevada commercial spent nuclear fuel transportation experience

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed.

  15. Nevada commercial spent nuclear fuel transportation experience

    International Nuclear Information System (INIS)

    1991-09-01

    The purpose of this report is to present an historic overview of commercial reactor spent nuclear fuel (SNF) shipments that have occurred in the state of Nevada, and to review the accident and incident experience for this type of shipments. Results show that between 1964 and 1990, 309 truck shipments covering approximately 40,000 miles moved through Nevada; this level of activity places Nevada tenth among the states in the number of truck shipments of SNF. For the same period, 15 rail shipments moving through the State covered approximately 6,500 miles, making Nevada 20th among the states in terms of number of rail shipments. None of these shipments had an accident or an incident associated with them. Because the data for Nevada are so limited, national data on SNF transportation and the safety of truck and rail transportation in general were also assessed

  16. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

  17. Basic study for development of nuclear heat application systems

    Energy Technology Data Exchange (ETDEWEB)

    Inaba, Yoshitomo; Fumizawa, Motoo; Hishida, Makoto [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1996-05-01

    We need to intensely investigate real possibilities of nuclear heat application systems which exploit high potential of nuclear energy as a promising candidate of the future energy resource in the world. In this report, special interest was placed on coal reforming systems because we thought a compact heat source of nuclear power with a very high energy density might compensate the environmental problem caused by burning a great amount of coal. First, we reviewed state-of-the-art technologies for coal reforming technology with a special attention on coal gasification technologies. Based on these basic data, we proposed several nuclear coal reforming systems and discussed advantages and disadvantages of the systems. We also explored a model with which we could analyze nuclear heat application systems all together. In addition, we investigated possibility and effects of nuclear heat utilization systems producing chemical materials from carbon dioxide in flue gas of fossil fuel power plant. As a result, we showed nuclear heat application systems were useful. (author).

  18. Transwaal - economic district heat from the Beznau nuclear power station

    International Nuclear Information System (INIS)

    Schatzmann, G.

    1986-01-01

    Initial study phases of the Transwaal project for distribution of heat from the Beznau nuclear power station via pipe lines to Aare and Limmat valley regions in Switzerland are presented. 500 MW heat availability through heat exchangers providing forward flow water temperature of 120 0 C, pipe line network and pumping station aspects, and the system energy flow diagram, are described. Considerations based on specific energy requirements in the year 2000 including alternative schemes showed economic viability. Investment and consumer costs and savings compared with oil and gas heating are discussed. Heat supply is guaranteed well into the 21st century and avoids environmental disadvantages. (H.V.H.)

  19. Nuclear transport factors: global regulation of mitosis.

    Science.gov (United States)

    Forbes, Douglass J; Travesa, Anna; Nord, Matthew S; Bernis, Cyril

    2015-08-01

    The unexpected repurposing of nuclear transport proteins from their function in interphase to an equally vital and very different set of functions in mitosis was very surprising. The multi-talented cast when first revealed included the import receptors, importin alpha and beta, the small regulatory GTPase RanGTP, and a subset of nuclear pore proteins. In this review, we report that recent years have revealed new discoveries in each area of this expanding story in vertebrates: (a) The cast of nuclear import receptors playing a role in mitotic spindle regulation has expanded: both transportin, a nuclear import receptor, and Crm1/Xpo1, an export receptor, are involved in different aspects of spindle assembly. Importin beta and transportin also regulate nuclear envelope and pore assembly. (b) The role of nucleoporins has grown to include recruiting the key microtubule nucleator - the γ-TuRC complex - and the exportin Crm1 to the mitotic kinetochores of humans. Together they nucleate microtubule formation from the kinetochores toward the centrosomes. (c) New research finds that the original importin beta/RanGTP team have been further co-opted by evolution to help regulate other cellular and organismal activities, ranging from the actual positioning of the spindle within the cell perimeter, to regulation of a newly discovered spindle microtubule branching activity, to regulation of the interaction of microtubule structures with specific actin structures. (d) Lastly, because of the multitudinous roles of karyopherins throughout the cell cycle, a recent large push toward testing their potential as chemotherapeutic targets has begun to yield burgeoning progress in the clinic. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. District heating by the Bohunice nuclear power plant

    International Nuclear Information System (INIS)

    Metke, E.; Skvarka, P.

    1984-01-01

    Technical and economical aspects of district heating by the electricity generating nuclear plants in Czechoslovakia are discussed. As a first stage of the project, 240 MW thermal power will be supplied using bleeding lines steam from the B-2 nuclear power plant at Jaslovske Bohunice to heat up water at a central station to 130 grad C. The maximal thermal power that can be produced for district heating by WWER type reactors with regular condensation turbines is estimated to be: 465 MW for a WWER-440 reactor with two 220 MWe turbines and 950 MW for a WWER-1000 reactor with a Skoda made 1000 MWe turbine using a three-stage scheme to heat up water from 60 grad C to 150 grad C. The use of satelite heating turbines connected to the steam collector is expected to improve the efficiency. District heating needs will de taken into account for siting of the new power plants

  1. Modeling transient heat transfer in nuclear waste repositories.

    Science.gov (United States)

    Yang, Shaw-Yang; Yeh, Hund-Der

    2009-09-30

    The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.

  2. Economics of long-distance transmission, storage, and distribution of heat from nuclear plants with existing and newer techniques

    International Nuclear Information System (INIS)

    Margen, P.H.

    1978-01-01

    Conventional and newer types of hot-water pipes are applied to the bulk transport of reject heat from central nuclear power plants to the district heating network of cities or groups of cities. With conventional pipes, the transport of 300 to 2000 MW of heat over distances of 30 to 100 km can be justified, while with newer pipe types, even longer distances would often be economic. For medium-size district heating schemes, low-temperature heat transport from simple heat-only reactors suitable for closer location to cities is of interest. For daily storage of heat on district heating systems, steel heat accumulators are currently used in Sweden. The development of more advanced cheaper heat accumulators, such as lake storage schemes, could make even seasonal heat storage economic. Newer distribution technology extends the economic field of penetration of district heating even to suburban one-family house districts. With proper design and optimization, nuclear district heating can be competitive in a wide market and achieve very substantial fossil-fuel savings

  3. Regulation on the transport of nuclear fuel materials by vehicles

    International Nuclear Information System (INIS)

    1984-01-01

    The regulations applying to the transport of nuclear fuel materials by vehicles, mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. The transport is for outside of the factories and the site of enterprises by such modes of transport as rail, trucks, etc. Covered are the following: definitions of terms, places of fuel materials handling, loading methods, limitations on mix loading with other cargo, radiation dose rates concerning the containers and the vehicles, transport indexes, signs and indications, limitations on train linkage during transport by rail, security guards, transport of empty containers, etc. together with ordinary rail cargo and so on. (Mori, K.)

  4. Heat transport in closed cell aluminum foams: application notes

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, Jaime; Escudero, Javier; Solorzano, Eusebio; Rodriguez-Perez, Miguel A.; Saja, Jose A. de [Cellular Materials Group (CellMat), Condensed Matter Physics Department, University of Valladolid (Spain)

    2009-10-15

    Heat transport equations have been used to solve, by implementing the Finite Element Method (FEM), three different cases representative of the aluminium foams life: the production process (solidification in the molten state), post-production (water quenching heat treatments) and applications (fire barriers). (Abstract Copyright [2009], Wiley Periodicals, Inc.)

  5. Heat transfer and mechanical interactions in fusion nuclear systems

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1984-01-01

    This general review of design issues in heat transfer and mechanical interactions of the first wall, blanket and shield systems of tokamak and mirror fusion reactors begins with a brief introduction to fusion nuclear systems. The design issues are summarized in tables and the following examples are described to illustrate these concerns: the surface heating of limiters, heat transfer from solid breeders, MHD effects in liquid metal blankets, mechanical loads from electromagnetic transients and remote maintenance

  6. Study of electronic heat transport in plasma through diagnosis based on modulated electron cyclotron heating; Etudes de transport de la chaleur electronique par injection modulee d'ondes a la frequence cyclotronique electronique

    Energy Technology Data Exchange (ETDEWEB)

    Clemencon, A.; Guivarch, C

    2003-07-01

    In order to make nuclear fusion energetically profitable, it is crucial to heat and confine the plasma efficiently. Studying the behavior of the heat diffusion coefficient is a key issue in this matter. The use of modulated electron cyclotron heating as a diagnostic has suggested the existence of a transport barrier under certain plasma conditions. We have determined the solution to the heat transport equation, for several heat diffusion coefficient profiles. By comparing the analytical solutions with experimental data; we are able to study the heat diffusion coefficient profile. Thus, in certain experiments, we can confirm that the heat diffusion coefficient switches from low to high values at the radius where the electron cyclotron heat deposition is made. (authors)

  7. Nuclear Energy and Synthetic Liquid Transportation Fuels

    Science.gov (United States)

    McDonald, Richard

    2012-10-01

    This talk will propose a plan to combine nuclear reactors with the Fischer-Tropsch (F-T) process to produce synthetic carbon-neutral liquid transportation fuels from sea water. These fuels can be formed from the hydrogen and carbon dioxide in sea water and will burn to water and carbon dioxide in a cycle powered by nuclear reactors. The F-T process was developed nearly 100 years ago as a method of synthesizing liquid fuels from coal. This process presently provides commercial liquid fuels in South Africa, Malaysia, and Qatar, mainly using natural gas as a feedstock. Nuclear energy can be used to separate water into hydrogen and oxygen as well as to extract carbon dioxide from sea water using ion exchange technology. The carbon dioxide and hydrogen react to form synthesis gas, the mixture needed at the beginning of the F-T process. Following further refining, the products, typically diesel and Jet-A, can use existing infrastructure and can power conventional engines with little or no modification. We can then use these carbon-neutral liquid fuels conveniently long into the future with few adverse environmental impacts.

  8. The eyes, ears and collective voice for nuclear transport

    International Nuclear Information System (INIS)

    Green, L.

    2000-01-01

    Transport is a vital part of the nuclear industry and the safety record of radioactive materials transport across the world is excellent. This record is due primarily to well-founded regulations developed by such intergovernmental organisations as the International Atomic Energy Agency and the International Maritime Organisation. It is due, also, to the professionalism of those in the industry. Attitudes to nuclear transport are important. They have the potential, if not heeded, and not responded to sensitively and convincingly to make life very much more difficult for those committed to the safe, reliable and efficient transport of nuclear materials. What is required is a balanced situation, which takes account both of the public's attitudes and industry's need for an efficient operation. The voices of the nuclear transport industry and those who value the industry need to be heard. The World Nuclear Transport Institute was established to provide the nuclear transport industry with the collective eyes, ears and voice in the key intergovernmental organisations which are so important to it. The nuclear transport industry has a safety record which could be regarded as a model for the transport of dangerous goods of all kinds. The industry is situated within a comprehensive and strict regime of national and international standards and regulations. That is the message to be disseminated, and that is the commitment of the World Nuclear Transport Institute as it works to protect and to promote the safe, efficient and reliable transport of radioactive materials. (author)

  9. The Thermos program for nuclear reactors specialized in district heating

    International Nuclear Information System (INIS)

    Lerouge, B.

    1976-01-01

    Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr

  10. Dedicated low temperature nuclear district heating plants: Rationale and prospects

    International Nuclear Information System (INIS)

    Goetzmann, C.A.

    1997-01-01

    Space heating accounts for a substantial fraction of the end-energy consumption in a large number of industrialized countries. Accordingly, efforts have been under way since many years to utilize nuclear energy as a source for district heating. The paper describes the key technical and institutional issues affecting the implementation of such technology. It is argued that the basic case for nuclear district heating is sound but that its introduction merits and drawbacks strongly depend on local circumstances. (author). 4 figs, 1 tab

  11. The Apatity nuclear heating plant project: modern technical and economic issues of nuclear heat application in Russia

    International Nuclear Information System (INIS)

    Adamov, E.O.; Romenkov, A.A.

    1998-01-01

    Traditionally Russia is a country with advanced structure of centralized heat supply. Many thermal plants and heating networks need technical upgrading to improve their technical and economic efficiency. Fossil fuelled heating capacities have a negative influence on ecology, which can be seen especially in the northern regions of Russia. Furthermore, fossil fuel prices are rising in Russia. The above factors tend to intensify the need for alternative heat sources being capable of solving the problem. Nuclear heat sources may be the alternative. In this paper, the main features of a proposed NHP in the Murmansk region are summarized. (author)

  12. An Overview of Liquid Fluoride Salt Heat Transport Systems

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL

    2010-09-01

    Heat transport is central to all thermal-based forms of electricity generation. The ever increasing demand for higher thermal efficiency necessitates power generation cycles transitioning to progressively higher temperatures. Similarly, the desire to provide direct thermal coupling between heat sources and higher temperature chemical processes provides the underlying incentive to move toward higher temperature heat transfer loops. As the system temperature rises, the available materials and technology choices become progressively more limited. Superficially, fluoride salts at {approx}700 C resemble water at room temperature being optically transparent and having similar heat capacity, roughly three times the viscosity, and about twice the density. Fluoride salts are a leading candidate heat-transport material at high temperatures. Fluoride salts have been extensively used in specialized industrial processes for decades, yet they have not entered widespread deployment for general heat transport purposes. This report does not provide an exhaustive screening of potential heat transfer media and other high temperature liquids such as alkali metal carbonate eutectics or chloride salts may have economic or technological advantages. A particular advantage of fluoride salts is that the technology for their use is relatively mature as they were extensively studied during the 1940s-1970s as part of the U.S. Atomic Energy Commission's program to develop molten salt reactors (MSRs). However, the instrumentation, components, and practices for use of fluoride salts are not yet developed sufficiently for commercial implementation. This report provides an overview of the current understanding of the technologies involved in liquid salt heat transport (LSHT) along with providing references to the more detailed primary information resources. Much of the information presented here derives from the earlier MSR program. However, technology has evolved over the intervening years

  13. Consequences of nonlinear heat transport laws on expected plasma profiles

    International Nuclear Information System (INIS)

    Lackner, K.

    1987-03-01

    The expected variation of plasma pressure profiles against changes in power deposition is investigated by using a simple linear heat transport law as well as a quadratic one. Applying the quadratic transport law it can be shown that the stiffening of the resulting profiles is sufficient to understand the experimentally measured phenomenon of 'profile consistence' without further assumptions of nonlocal effects. (orig.) [de

  14. Multipurpose nuclear process heat for energy supply in Brazil

    International Nuclear Information System (INIS)

    Hansen, U.; Inden, P.; Oesterwind, D.; Hukai, R.Y.; Pessine, R.T.; Pieroni, R.R.; Visoni, E.

    1978-11-01

    The industrialized nations require 75% of the energy as heat and it is likely that developing countries in the course of industrialization will show a comparable energy consumption structure. The High Temperature Reactor (HTR) allows the utilization of nuclear energy at high temperatures as process heat. In the Federal Republic of Germany (FRG) the development in the relevant technical areas is well advanced and warrants investigation as a matter for transfer to Brazil. In Brazil nuclear process heat finds possible applications in steel making, shale oil extraction, petroleum refining, and in the more distant future coal gasification with distribution networks. Based on growth forecasts for these industries a theoretical potential market of 38-53 GW (th) can be identified. At present nuclear process heat is marginally more expensive than conventional fossil technologies but the anticipated development is expected to add an economic incentive to the emerging necessity of providing a sound energy base in the developing countries. (author)

  15. Small reactors for low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    1988-06-01

    In accordance with the Member States' calls for information exchange in the field of nuclear heat application (NHA) two IAEA meetings were organized already in 1976 and 1977. After this ''promising period'', the development of relevant programmes in IAEA Member States was slowed down and therefore only after several years interruption a new Technical Committee Meeting with a Workshop was organized in late 1983, to review the status of NHA, after a few new specific plans appeared in some IAEA Member States in the early 1980's for the use of heat from existing or constructed NPPs and for developing nuclear heating plants (NHP). In June 1987 an Advisory Group Meeting was convened in Winnipeg, Canada, to discuss and formulate a state-of-the-art review on ''Small Reactors for Low Temperature Nuclear Heat Application''. Information on this subject gained up to 1987 in the Member States whose experts attended this meeting is embodied in the present Technical Report. Figs and tabs

  16. Heat and Moisture transport of socks

    Science.gov (United States)

    Komárková, P.; Glombíková, V.; Havelka, A.

    2017-10-01

    Investigating the liquid moisture transport and thermal properties is essential for understanding physiological comfort of clothes. This study reports on an experimental investigation of moisture management transport and thermal transport on the physiological comfort of commercially available socks. There are subjective evaluation and objective measurements. Subjective evaluation of the physiological comfort of socks is based on individual sensory perception of probands during and after physical exertion. Objective measurements were performed according to standardized methods using Moisture Management tester for measuring the humidity parameters and C-term TCi analyzer for thermal conductivity and thermal effusivity. The obtained values of liquid moisture transport and thermal properties were related to the material composition and structure of the tested socks. In summary, these results show that objective measurement corresponds with probands feelings.

  17. Materials for nuclear diffusion-bonded compact heat exchangers

    International Nuclear Information System (INIS)

    Li, Xiuqing; Smith, Tim; Kininmont, David; Dewson, Stephen John

    2009-01-01

    This paper discusses the characteristics of materials used in the manufacture of diffusion bonded compact heat exchangers. Heatric have successfully developed a wide range of alloys tailored to meet process and customer requirements. This paper will focus on two materials of interest to the nuclear industry: dual certified SS316/316L stainless steel and nickel-based alloy Inconel 617. Dual certified SS316/316L is the alloy used most widely in the manufacture of Heatric's compact heat exchangers. Its excellent mechanical and corrosion resistance properties make it a good choice for use with many heat transfer media, including water, carbon dioxide, liquid sodium, and helium. As part of Heatric's continuing product development programme, work has been done to investigate strengthening mechanisms of the alloy; this paper will focus in particular on the effects of nitrogen addition. Another area of Heatric's programme is Alloy 617. This alloy has recently been developed for diffusion bonded compact heat exchanger for high temperature nuclear applications, such as the intermediate heat exchanger (IHX) for the very high temperature nuclear reactors for production of electricity, hydrogen and process heat. This paper will focus on the effects of diffusion bonding process and cooling rate on the properties of alloy 617. This paper also compares the properties and discusses the applications of these two alloys to compact heat exchangers for various nuclear processes. (author)

  18. Potential of low-temperature nuclear heat applications

    International Nuclear Information System (INIS)

    1986-12-01

    At present, more than one third of the fossil fuel currently used is being consumed to produce space heating and to meet industrial needs in many countries of the world. Imported oil still represents a large portion of this fossil fuel and despite its present relatively low price future market evolutions with consequent upward cost revisions cannot be excluded. Thus the displacement of the fossil fuel by cheaper low-temperature heat produced in nuclear power plants is a matter which deserves careful consideration. Technico-economic studies in many countries have shown that the use of nuclear heat is fully competitive with most of fossil-fuelled plants, the higher investment costs being offset by lower production cost. Another point in favour of heat generation by nuclear source is its indisputable advantage in terms of benefits to the environment. The IAEA activity plans for 1985-86 concentrate on information exchange with specific emphasis on the design criteria, operating experience, safety requirements and specifications of heat-only reactors, co-generation plants and existing power plants backfitted for additional heat applications. The information gained up to 1985 was discussed during the Advisory Group Meeting on the Potential of Low-Temperature Nuclear Heat Applications held in the Federal Institute for Reactor Research, Wuerenlingen, Switzerland in September 1985 and, is included in the present Technical Document

  19. Perspectives of heat transfer enhancement in nuclear reactors toward nanofluids applications

    International Nuclear Information System (INIS)

    Rocha, Marcelo S.; Cabral, Eduardo L.L.; Sabundjian, Gaiane

    2013-01-01

    Nanofluids are colloidal suspensions of nanoparticles in a base fluid with interesting physical properties and large potential for heat transfer enhancement in thermal systems among other applications. There are an increasing number of nanofluids investigations concerning many aspects of synthesis and fabrication technologies, physical properties, and special applications. Results demonstrate that physical properties like high thermal conductivities and high critical heat flux (CHF) of some nanofluids classifies them as potential working fluids for high heat flux transportation in special systems, including thermal management of microelectronic devices (MEMS) and nuclear reactors. Understanding the importance of such investigations for the knowledge development of nuclear engineering a new research is being conducted at the Nuclear Engineering Center (CEN) of the Nuclear and Energy Research Institute (IPEN/CNEN-SP) to analyze the application potentiality of some nanofluids in nuclear systems for heat transfer enhancement under ionizing radiation influence. In this work a revision of theoretical and experimental studies of nanofluids is performed and its potentiality for using in future generations of nuclear reactors is highlighted showing the status of the research at present. (author)

  20. High efficiency heat transport and power conversion system for cascade

    International Nuclear Information System (INIS)

    Maya, I.; Bourque, R.F.; Creedon, R.L.; Schultz, K.R.

    1985-02-01

    The Cascade ICF reactor features a flowing blanket of solid BeO and LiAlO 2 granules with very high temperature capability (up to approx. 2300 K). The authors present here the design of a high temperature granule transport and heat exchange system, and two options for high efficiency power conversion. The centrifugal-throw transport system uses the peripheral speed imparted to the granules by the rotating chamber to effect granule transport and requires no additional equipment. The heat exchanger design is a vacuum heat transfer concept utilizing gravity-induced flow of the granules over ceramic heat exchange surfaces. A reference Brayton power cycle is presented which achieves 55% net efficiency with 1300 K peak helium temperature. A modified Field steam cycle (a hybrid Rankine/Brayton cycle) is presented as an alternate which achieves 56% net efficiency

  1. Phonon hydrodynamics for nanoscale heat transport at ordinary temperatures

    Science.gov (United States)

    Guo, Yangyu; Wang, Moran

    2018-01-01

    The classical Fourier's law fails in extremely small and ultrafast heat conduction even at ordinary temperatures due to strong thermodynamic nonequilibrium effects. In this work, a macroscopic phonon hydrodynamic equation beyond Fourier's law with a relaxation term and nonlocal terms is derived through a perturbation expansion to the phonon Boltzmann equation around a four-moment nonequilibrium solution. The temperature jump and heat flux tangential retardant boundary conditions are developed based on the Maxwell model of the phonon-boundary interaction. Extensive steady-state and transient nanoscale heat transport cases are modeled by the phonon hydrodynamic model, which produces quantitative predictions in good agreement with available phonon Boltzmann equation solutions and experimental results. The phonon hydrodynamic model provides a simple and elegant mathematical description of non-Fourier heat conduction with a clear and intuitive physical picture. The present work will promote deeper understanding and macroscopic modeling of heat transport in extreme states.

  2. Economic Analyses and Potential Market of the 200MW Nuclear Heating Reactor

    International Nuclear Information System (INIS)

    Wang, Yongqing; Wang, Guiying

    1992-01-01

    Based on the 5MW experimental nuclear heating reactor, Intent has developed a 200MW demonstration nuclear heating reactor. Owing to its simplified systems and low operating parameters, the NCR-200 has preferable investment in comparison with that of a nuclear power plant. The pre-feasibility studies for several cities in Northern China have shown that the heat cost of a NCR-200 can be competitive with a coal fired heating plant. As a safe, clean and economic heat source, the NCR could pose a large market in replacement of coal for heating. The R and D work performed up to now has demonstrated that the NCR-200 operating under the present parameters can supply low pressure steam for industrial process and co-generation to enhance it economic benefit. The NCR-200 could also serve a heat source for air condition by using Li Br refrigerator, this application is very interesting to some cities in Central and Southern China. The applications of the NCR in oil recovery by injecting hot water and transportation are very promising for some oil fields in North China. In addition, the study on sea water desalination using the NCR-200 is being carried out at present under international cooperation. All of these will expansion the possible application of the NCR. The paper presents the economic analysis and the potential market of the NCR-200

  3. Liability and insurance aspects of international transport of nuclear materials

    International Nuclear Information System (INIS)

    van Gijn, S.H.

    1985-01-01

    The Paris and Vienna Conventions do not affect the application of any international transport agreement already in force. However, in certain circumstances both the nuclear operator and the carrier may be held liable for nuclear damage which arises during international transports of nuclear materials. The ensuing cumulation of liabilities under the Nuclear and Transport Conventions may cause serious problems in obtaining adequate insurance cover for such transports. The 1971 Brussels Convention seeks to solve this problem by exonerating any person who might be held liable for nuclear damage under an international maritime convention or national law. Similar difficulties are encountered in the case of transports of nuclear materials between states which have and states which have not ratified the Paris and Vienna Conventions. (NEA) [fr

  4. The molecular mechanism for nuclear transport and its application.

    Science.gov (United States)

    Kim, Yun Hak; Han, Myoung-Eun; Oh, Sae-Ock

    2017-06-01

    Transportation between the cytoplasm and the nucleoplasm is critical for many physiological and pathophysiological processes including gene expression, signal transduction, and oncogenesis. So, the molecular mechanism for the transportation needs to be studied not only to understand cell physiological processes but also to develop new diagnostic and therapeutic targets. Recent progress in the research of the nuclear transportation (import and export) via nuclear pore complex and four important factors affecting nuclear transport (nucleoporins, Ran, karyopherins, and nuclear localization signals/nuclear export signals) will be discussed. Moreover, the clinical significance of nuclear transport and its application will be reviewed. This review will provide some critical insight for the molecular design of therapeutics which need to be targeted inside the nucleus.

  5. Heat transfer challenges in the nuclear industry

    International Nuclear Information System (INIS)

    Penot, F.; Petit, D.; El Ganaoui, M.; Fauchais, P.

    2006-01-01

    The 2006 issue of the annual French congress of thermal engineering has chosen as specific topic the nuclear industry. Nuclear fission represents today about 12% of the energy production and more than 80% of the power generation in France. Its use at the worldwide scale is less important but the economic growth of countries like India and China would deeply change this situation. Different aspects of nuclear industry have been approached during the congress: the worldwide development of nuclear energy in a market environment, the problems linked with the geologic disposal of high level and long lived radioactive wastes, the stakes and thermal challenges of present and future nuclear systems, the R and D challenges of future systems and the general problem of tomorrow's energy supplies. The domain of very high temperatures, largely present in future systems and the domain of thermonuclear fusion have been also approached. Twenty-two papers out of 192 presented at the congress have been selected which fit with the INIS scope. (J.S.)

  6. Phonon heat transport in gallium arsenide

    Indian Academy of Sciences (India)

    The energy linewidth is found to be an extremely sensitive quantity in the transport phenomena of crystalline solids as a collection of large number of scattering processes, namely, boundary scattering, impurity scattering, multiphonon scattering, interference scattering, electron–phonon processes and resonance scattering.

  7. Desalination by very low temperature nuclear heat

    International Nuclear Information System (INIS)

    Saari, Risto

    1977-01-01

    A new sea water desalination method has been developed: Nord-Aqua Vacuum Evaporation, which utilizes waste heat at a very low temperature. The requisite vacuum is obtained by the aid of a barometric column and siphon, and the dissolved air is removed from the vacuum by means of water flows. According to test results from a pilot plant, the process is operable if the waste heat exists at a temperature 7degC higher than ambient. The pumping energy which is then required is 9 kcal/kg, or 1.5% of the heat of vaporization of water. Calculations reveal that the method is economically considerably superior to conventional distilling methods. (author)

  8. Force Triggers YAP Nuclear Entry by Regulating Transport across Nuclear Pores.

    Science.gov (United States)

    Elosegui-Artola, Alberto; Andreu, Ion; Beedle, Amy E M; Lezamiz, Ainhoa; Uroz, Marina; Kosmalska, Anita J; Oria, Roger; Kechagia, Jenny Z; Rico-Lastres, Palma; Le Roux, Anabel-Lise; Shanahan, Catherine M; Trepat, Xavier; Navajas, Daniel; Garcia-Manyes, Sergi; Roca-Cusachs, Pere

    2017-11-30

    YAP is a mechanosensitive transcriptional activator with a critical role in cancer, regeneration, and organ size control. Here, we show that force applied to the nucleus directly drives YAP nuclear translocation by decreasing the mechanical restriction of nuclear pores to molecular transport. Exposure to a stiff environment leads cells to establish a mechanical connection between the nucleus and the cytoskeleton, allowing forces exerted through focal adhesions to reach the nucleus. Force transmission then leads to nuclear flattening, which stretches nuclear pores, reduces their mechanical resistance to molecular transport, and increases YAP nuclear import. The restriction to transport is further regulated by the mechanical stability of the transported protein, which determines both active nuclear transport of YAP and passive transport of small proteins. Our results unveil a mechanosensing mechanism mediated directly by nuclear pores, demonstrated for YAP but with potential general applicability in transcriptional regulation. Copyright © 2017 Elsevier Inc. All rights reserved.

  9. Development of heat exchangers for nuclear service

    International Nuclear Information System (INIS)

    Hodge, R.I.; Dalrymple, D.G.

    1976-01-01

    Unusual design constraints, due to tube vibration, are called for when tube-in-shell heat exchangers are incorporated into CANDU type reactor power plants. CRNL has programs studying tube excitation and response, flow conditions, and the fretting process in such exchangers, tube plugging techniques, and eddy current scanning systems for inside bores of full-length tubes. (E.C.B.)

  10. Passive heat transport in advanced CANDU containment

    International Nuclear Information System (INIS)

    Krause, M.; Mathew, P.M.

    1994-01-01

    A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

  11. Ordinance concerning the filing of transport of nuclear fuel materials

    International Nuclear Information System (INIS)

    1987-01-01

    This Order provides provisions concerning nuclear fuel substances requiring notification (nuclear fuel substance, material contaminated with nuclear fuel substances, fissionable substances, etc.), procedure for notification (to prefectural public safety commission), certificate of transpot (issued via public safety commission), instructions (speed of vehicle for transporting nuclear fuel substances, parking of vehicle, place for loading and unloading of nuclear fuel substances, method for loading and unloading, report to police, measures for disaster prevention during transport, etc.), communication among members of public safety commission (for smooth transport), notification of alteration of data in transport certificate (application to be submitted to public safety commission), application of reissue of transport certificate, return of transport certificate, inspection concerning transport (to be performed by police), submission of report (to be submitted by refining facilities manager, processing facilities manager, nuclear reactor manager, master of foreign nuclear powered ship, reprocessing facilities manager, waste disposal facilities manager; concerning stolen or missing nuclear fuel substances, traffic accident, unusual leakage of nuclear fuel substances, etc.). (Nogami, K.)

  12. Seawater desalination using nuclear heat/electricity - Prospects and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Misra, B.M. [Division of Nuclear Power, International Atomic Energy Agency (IAEA), Vienna (Austria)

    2007-02-15

    The desalination of seawater using nuclear energy is a feasible option to meet the growing demand of potable water. Over 150 reactor-years of operating experience of nuclear desalination have been accumulated worldwide. Several demonstration programs of nuclear desalination are also in progress to confirm its technical and economic viability under country specific conditions, with technical co-ordination or support of IAEA. Recent techno-economic feasibility studies carried out by some Member States indicate the competitiveness of nuclear desalination. This paper presents the salient activities on nuclear desalination in the Agency and in the interested Member States. In particular, some data from the feasibility studies carried out in the two south Mediterranean countries - Egypt and Tunisia - for their nuclear power and desalination projects are presented. Economic research on further water cost reduction includes investigation on utilization of waste heat from different nuclear reactor types for thermal desalination, preheat reverse osmosis using condenser cooling water return as feed and hybrid MED-RO desalination systems. The details of these are discussed. The main challenge for the large-scale deployment of nuclear seawater desalination is the lack of infrastructure and resources in the countries affected by water scarcity problems which are, however, interested in adoption of nuclear desalination for the sustainable water resources in the coming years. Socioeconomic and environmental aspects and the public perception for the nuclear desalination projects are also important factors requiring greater information exchange. These aspects are discussed while considering an integrated nuclear desalination system. (author)

  13. Choice of flowsheet for heat storage plant of nuclear stations for distant heat and power supply

    International Nuclear Information System (INIS)

    Larin, E.A.; Lut'yanov, A.F.

    1986-01-01

    A rational utilization flowsheet of converted gas heat in heat-accumulating part (HAP) for NPP distant heat supply (NPDH), in which high-temperature helium heat of HTGR type reactor is accumulated in chemically bound state in the process of methane steam conversion, and then it is transported for long distances, is suggested. Technical and economical analysis has shown, that the NPDH HAP most effective flowsheet envisages the use of dry saturated steam, generated at the expense of converted gas heat, as a coolant for intermediate turbine superheater. The high-temperature heat can be more effectively used for gaseous heating of supply water and basic condensate of the turbine installation, and low-temperature one - for heat supply of the near-located users

  14. National need for utilizing nuclear energy for process heat generation

    International Nuclear Information System (INIS)

    Gambill, W.R.; Kasten, P.R.

    1984-01-01

    Nuclear reactors are potential sources for generating process heat, and their applications for such use economically competitive. They help satisfy national needs by helping conserve and extend oil and natural gas resources, thus reducing energy imports and easing future international energy concerns. Several reactor types can be utilized for generating nuclear process heat; those considered here are light water reactors (LWRs), heavy water reactors (HWRs), gas-cooled reactors (GCRs), and liquid metal reactors (LMRs). LWRs and HWRs can generate process heat up to 280 0 C, LMRs up to 540 0 C, and GCRs up to 950 0 C. Based on the studies considered here, the estimated process heat markets and the associated energy markets which would be supplied by the various reactor types are summarized

  15. Experience of air transport of nuclear fuel material in Japan

    International Nuclear Information System (INIS)

    Yamashita, T.; Toguri, D.; Kawasaki, M.

    2004-01-01

    Certified Reference Materials (hereafter called as to CRMs), which are indispensable for Quality Assurance and Material Accountability in nuclear fuel plants, are being provided by overseas suppliers to Japanese nuclear entities as Type A package (non-fissile) through air transport. However, after the criticality accident at JCO in Japan, special law defining nuclear disaster countermeasures (hereafter called as to the LAW) has been newly enforced in June 2000. Thereafter, nuclear fuel materials must meet not only to the existing transport regulations but also to the LAW for its transport

  16. Generalized heat-transport equations: parabolic and hyperbolic models

    Science.gov (United States)

    Rogolino, Patrizia; Kovács, Robert; Ván, Peter; Cimmelli, Vito Antonio

    2018-03-01

    We derive two different generalized heat-transport equations: the most general one, of the first order in time and second order in space, encompasses some well-known heat equations and describes the hyperbolic regime in the absence of nonlocal effects. Another, less general, of the second order in time and fourth order in space, is able to describe hyperbolic heat conduction also in the presence of nonlocal effects. We investigate the thermodynamic compatibility of both models by applying some generalizations of the classical Liu and Coleman-Noll procedures. In both cases, constitutive equations for the entropy and for the entropy flux are obtained. For the second model, we consider a heat-transport equation which includes nonlocal terms and study the resulting set of balance laws, proving that the corresponding thermal perturbations propagate with finite speed.

  17. Slowpoke: a role for nuclear technology in district heating

    International Nuclear Information System (INIS)

    Lynch, G.F.

    1987-08-01

    The successful application of the SLOWPOKE concept to satisfy the heating needs of institutions and building complexes is described. Although the load factor for heating in Japan may not be as high as those experienced in other countries of the northern hemipshere, this particular application clearly demonstrates that small, special purpose, ultra-safe nuclear energy sources are technically and economically viable. They can be designed for easy operation and maintenance, to be located in urban areas and remote communities, thereby satsifying a broad spectrum of energy needs that cannot be served by central nuclear electrical generators

  18. Public and media acceptance of nuclear materials transport

    International Nuclear Information System (INIS)

    Lindeman, E.

    1999-01-01

    Transport is absolutely essential to the continued existence of a nuclear industry that includes large-scale power generation, sophisticated research, and medicine. Indeed, transport of nuclear materials is hardly a new business. What is new is the public's awareness and distrust of this transport - a distrust fuelled by the well-funded and skilled manipulation of the nuclear industry's detractors. The nuclear industry itself has only recently begun to acknowledge the importance and the implications of transport. This paper looks at the public and media response to the European-Japanese and the US Department of Energy's transport campaigns and quotes from several telling newspaper articles. It emphasizes the need for the nuclear industry to continue to be vigilant in its efforts to reach the public, media and governments with good science, openness and well-communicated facts. (author)

  19. Use of waste heat from nuclear power plants

    International Nuclear Information System (INIS)

    Olszewski, M.

    1978-01-01

    The paper details the Department of Energy (DOE) program concerning utilization of power plant reject heat conducted by the Oak Ridge National Laboratory (ORNL). A brief description of the historical development of the program is given and results of recent studies are outlined to indicate the scope of present efforts. A description of a DOE-sponsored project assessing uses for reject heat from the Vermont Yankee Nuclear Station is also given

  20. Heating of water by nuclear power stations

    International Nuclear Information System (INIS)

    Many studies have been carried out both in France and abroad to determine the magnitude of the thermal effects due to cooling of conventional or nuclear open-circuit electricity-generating stations. After an account of the main effects observed and the information derived from these results the discussion centres on the conclusions drawn from American research by the two bodies responsible for checking the impact of nuclear stations on the environment in the United States. The possible effects on aquatic organisms exposed to hot water discharges are of two kinds: the direct thermal effects (changes in the metabolism, growth rate and period of sexual maturity increase in mortality, etc..) and the indirect thermal effects (asphyxia due to a drop in the dissolved oxygen concentration, changes in the sensitivity to certain toxic substances, in resistance to infection, in the food chain) [fr

  1. Heat pipe heat transport system for the Stirling Space Power Converter (SSPC)

    Science.gov (United States)

    Alger, Donald L.

    1992-08-01

    Life issues relating to a sodium heat pipe heat transport system are described. The heat pipe system provides heat, at a temperature of 1050 K, to a 50 kWe Stirling engine/linear alternator power converter called the Stirling Space Power Converter (SSPC). The converter is being developed under a National Aeronautics and Space Administration program. Since corrosion of heat pipe materials in contact with sodium can impact the life of the heat pipe, a literature review of sodium corrosion processes was performed. It was found that the impurity reactions, primarily oxygen, and dissolution of alloy elements were the two corrosion process likely to be operative in the heat pipe. Approaches that are being taken to minimize these corrosion processes are discussed.

  2. Japan's regulatory and safety issues regarding nuclear materials transport

    International Nuclear Information System (INIS)

    Saito, T.; Yamanaka, T.

    2004-01-01

    This paper focuses on the regulatory and safety issues on nuclear materials transport which the Government of Japan (GOJ) faces and needs to well handle. Background information about the status of nuclear power plants (NPP) and nuclear fuel cycle (NFC) facilities in Japan will promote a better understanding of what this paper addresses

  3. Thermodynamic framework for a generalized heat transport equation

    Directory of Open Access Journals (Sweden)

    Guo Yangyu

    2016-06-01

    Full Text Available In this paper, a generalized heat transport equation including relaxational, nonlocal and nonlinear effects is provided, which contains diverse previous phenomenological models as particular cases. The aim of the present work is to establish an extended irreversible thermodynamic framework, with generalized expressions of entropy and entropy flux. Nonlinear thermodynamic force-flux relation is proposed as an extension of the usual linear one, giving rise to the nonlinear terms in the heat transport equation and ensuring compatibility with the second law. Several previous results are recovered in the linear case, and some additional results related to nonlinear terms are also obtained.

  4. Heating control system for nuclear reactor

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1981-01-01

    Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

  5. Study on neutron diffusion and time dependence heat ina fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Vilhena, M.T. de.

    1988-01-01

    The purpose of this work is to model the neutron diffusion and heat transfer for a Fluidized Bed Nuclear Reactor and its solution by Laplace Transform Technique with numerical inversion using Fourier Series. Also Gaussian quadrature and residues techniques were applied for numerical inversion. The neutron transport, diffusion, and point Kinetic equation for this nuclear reactor concept are developed. A matricial and Taylor Series methods are proposed for the solution of the point Kinetic equation which is a time scale problem of Stiff type

  6. Calculating iron transport in nuclear systems

    International Nuclear Information System (INIS)

    Horowitz, J.S.; Merilo, M.; Munson, D.

    2002-01-01

    The presence of high levels of iron in the final feedwater of nuclear plants is undesirable and can have a significant contribution to plant operations and maintenance (O and M) costs. A number of options are available to reduce the iron concentration, but tend to be expensive. Recently a method was developed to quantitatively determine the contribution of each iron source, such that reduction options can be quantitatively compared. The method is based on industry experience that the majority of iron has been released by flow-accelerated corrosion (FAC). FAC is one of the most predictable forms of corrosion and a well-developed predictive model has been developed and also encoded in the CHECWORKS. A combination of CHECWORKS and supplemental calculations have been used to model the iron transport in a number of US BWRs and PWRs. The iron generated by FAC in all the normally operating piping systems has been calculated using the results of CHECWORKS predictions and a special post processor. The post processor accounts for the differences between the maximum corrosion rate calculated by CHECWORKS and the average corrosion (iron generation) rate for a pipe-fitting or length of pipe. It also calculates the amount of iron generated within the fitting or pipe. Supplemental calculations have been used to determine the iron generation from the major, in-line components - high and low pressure turbines, moisture separators, feedwater heaters and the condenser. All of the iron generation rates for the equipment and piping were appropriately summed and iron concentrations estimated throughout the steam-feedwater system. Predicted iron concentrations have agreed well with plant measurements. The availability of specific iron generation rates allows plant management to make reasoned decisions about the countermeasures to deal with iron generation and transport. The countermeasures that have been examined to reduce the amount of iron transport include installing additional water

  7. Phonon hydrodynamics and its applications in nanoscale heat transport

    Science.gov (United States)

    Guo, Yangyu; Wang, Moran

    2015-09-01

    Phonon hydrodynamics is an effective macroscopic method to study heat transport in dielectric solid and semiconductor. It has a clear and intuitive physical picture, transforming the abstract and ambiguous heat transport process into a concrete and evident process of phonon gas flow. Furthermore, with the aid of the abundant models and methods developed in classical hydrodynamics, phonon hydrodynamics becomes much easier to implement in comparison to the current popular approaches based on the first-principle method and kinetic theories involving complicated computations. Therefore, it is a promising tool for studying micro- and nanoscale heat transport in rapidly developing micro and nano science and technology. However, there still lacks a comprehensive account of the theoretical foundations, development and implementation of this approach. This work represents such an attempt in providing a full landscape, from physical fundamental and kinetic theory of phonons to phonon hydrodynamics in view of descriptions of phonon systems at microscopic, mesoscopic and macroscopic levels. Thus a systematical kinetic framework, summing up so far scattered theoretical models and methods in phonon hydrodynamics as individual cases, is established through a frame of a Chapman-Enskog solution to phonon Boltzmann equation. Then the basic tenets and procedures in implementing phonon hydrodynamics in nanoscale heat transport are presented through a review of its recent wide applications in modeling thermal transport properties of nanostructures. Finally, we discuss some pending questions and perspectives highlighted by a novel concept of generalized phonon hydrodynamics and possible applications in micro/nano phononics, which will shed more light on more profound understanding and credible applications of this new approach in micro- and nanoscale heat transport science.

  8. Transport and storage of spent nuclear fuel

    International Nuclear Information System (INIS)

    Lung, M.; Lenail, B.

    1987-01-01

    From a safety standpoint, spent fuel is clearly not ideal for permanent disposal and reprocessing is the best method of preparing wastes for long-term storage in a repository. Furthermore, the future may demonstrate that some fission products recovered in reprocessing have economic applications. Many countries have in fact reached the point at which the recycling of plutonium and uranium from spent fuel is economical in LWR's. Even in countries where this is not yet evident, (i.e., the United States), the French example shows that the day will come when spent fuel will be retrieved for reprocessing and recycle. It is highly questionable whether spent fuel will ever be considered and treated as waste in the same sense as fission products and processed as such, i.e., packaged in a waste form for permanent disposal. Even when recycled fuel material can no longer be reused in LWR's because of poor reactivity, it will be usable in FBR's. Based on the considerable experience gained by SGN and Cogema, this paper has provided practical discussion and illustrations of spent fuel transport and storage of a very important step in the nuclear fuel management process. The best of spent fuel storage depends on technical, economic and policy considerations. Each design has a role to play and we hope that the above discussion will help clarify certain issues

  9. [Worker heat disorders at the Fukushima Daiichi nuclear power plant].

    Science.gov (United States)

    Tsuji, Masayoshi; Kakamu, Takeyasu; Hayakawa, Takehito; Kumagai, Tomohiro; Hidaka, Tomoo; Kanda, Hideyuki; Fukushima, Tetsuhito

    2013-01-01

    Ever since the Fukushima Daiichi nuclear power plant (NPP) accident, every day about 3,000 workers have been working to repair the situation. The frequent occurrence of heat disorders has been a concern for the workers wearing protective clothing with poor ventilation. We have been analyzing the heat disorder problem since the accident in order to come up with a solution to prevent future heat disorder incidents among Fukushima Daiichi NPP accident clean-up workers. From March 22 to September 16, 2011, the Fukushima Labor Bureau assessed 43 cases of nuclear power plant workers with heat disorders. Age of subject, month and time of occurrence, temperature, and humidity were examined for each case, as well as the severity of heat disorders. The grade of severity was divided into Grade I and Grade II or higher. Then, age, temperature, and humidity were analyzed using the Mann-Whitney Utest, and age, temperature, humidity, and presence or absence of a cool-vest were analyzed using the χ(2) test and logistic regression analysis. SPSS version 17.0 statistical software was used with a level of significance of pFukushima Daiichi NPP occurred in clean-up workers at the relatively younger ages of 30-40, suggesting the need for heat disorder prevention measures for these younger workers. Heat disorder cases primarily occurred in the morning, necessitating preventive measures for the early hours of the day. In addition, because heat disorders of Grade II or higher occurred in June in 5 of 10 cases, we believe heat disorder precautions should be implemented from June. The lack of significant difference in severity difference may be attributable to the small number of cases or other factors. We think Fukushima Daiichi NPP accident clean-up workers need heat disorder prevention measures for their safety, based on the results of this study.

  10. Ultimate after-heat removal system for nuclear reactors

    International Nuclear Information System (INIS)

    Bernard, L. Jr.

    1980-01-01

    The invention concerns the safety region of a nuclear power plant, especially the divertor for the residual heat which keeps forming after shutdown of the reactor. According to the invention a dry cooling tower of enclosed construction is planned. The walls and roof shall be rocket-proof. Such a configuration is described and explained by means of designs. (UWI) [de

  11. Valve arrangement for a nuclear plant residual heat removal system

    International Nuclear Information System (INIS)

    Fidler, G.L.; Hill, R.A.; Carrera, J.P.

    1978-01-01

    Disclosed is an improved valve arrangement for a two-train Residual Heat Removal System (RHRS) of a nuclear reactor plant which ensures operational integrity of the system under single failure circumstances including loss of one of two electrical power sources

  12. Magnetic-field asymmetry of nonlinear thermoelectric and heat transport

    International Nuclear Information System (INIS)

    Hwang, Sun-Yong; Sánchez, David; López, Rosa; Lee, Minchul

    2013-01-01

    Nonlinear transport coefficients do not obey, in general, reciprocity relations. We here discuss the magnetic-field asymmetries that arise in thermoelectric and heat transport of mesoscopic systems. Based on a scattering theory of weakly nonlinear transport, we analyze the leading-order symmetry parameters in terms of the screening potential response to either voltage or temperature shifts. We apply our general results to a quantum Hall antidot system. Interestingly, we find that certain symmetry parameters show a dependence on the measurement configuration. (paper)

  13. FFTF Heat Transport System (HTS) component and system design

    International Nuclear Information System (INIS)

    Young, M.W.; Edwards, P.A.

    1980-01-01

    The FFTF Heat Transport Systems and Components designs have been completed and successfully tested at isothermal conditions up to 427 0 C (800 0 F). General performance has been as predicted in the design analyses. Operational flexibility and reliability have been outstanding throughout the test program. The components and systems have been demonstrated ready to support reactor powered operation testing planned later in 1980

  14. Desalination demonstration plant using nuclear heat

    International Nuclear Information System (INIS)

    Hanra, M.S.; Misra, B.M.

    1998-01-01

    Most of the desalination plants which are operating throughout the world utilize the energy from thermal power station which has the main disadvantage of polluting the environment due to combustion of fossil fuel and with the inevitable rise in prices of fossil fuel, nuclear driven desalination plants will become more economical. So it is proposed to set up nuclear desalination demonstration plant at the location of Madras Atomic Power Station (MAPS), Kalpakkam. The desalination plant will be of a capacity 6300 m 3 /day and based on both Multi Stage Flash (MSF) and Sea Water Reverse Osmosis (SWRO) processes. The MSF plant with performance ratio of 9 will produce water total dissolved solids (TDS-25 ppm) at a rate of 4500 m 3 /day from seawater of 35000 ppm. A part of this water namely 1000 m 3 /day will be used as Demineralised (DM) water after passing it through a mixed bed polishing unit. The remaining 3500 m 3 /day water will be mixed with 1800 m 3 /day water produced from the SWRO plant of TDS of 400 ppm and the same be supplied to industrial/municipal use. The sea water required for MSF and SWRO plants will be drawn from the intake/outfall system of MAPS which will also supply the required electric power pumping. There will be net 4 MW loss of power of MAPS namely 3 MW for MSF and 1 MW for SWRO desalination plants. The salient features of the project as well as the technical details of the both MSF and SWRO processes and its present status are given in this paper. It also contains comparative cost parameters of water produced by both processes. (author)

  15. Two-phase flow heat transfer in nuclear reactor systems

    International Nuclear Information System (INIS)

    Koncar, Bostjan; Krepper, Eckhard; Bestion, Dominique; Song, Chul-Hwa; Hassan, Yassin A.

    2013-01-01

    Complete text of publication follows: Heat transfer and phase change phenomena in two-phase flows are often encountered in nuclear reactor systems and are therefore of paramount importance for their optimal design and safe operation.The complex phenomena observed especially during transient operation of nuclear reactor systems necessitate extensive theoretical and experimental investigations. This special issue brings seven research articles of high quality. Though small in number, they cover a wide range of topics, presenting high complexity and diversity of heat transfer phenomena in two-phase flow. In the last decades a vast amount of research has been devoted to theoretical work and computational simulations, yet the experimental work remains indispensable for understanding of two-phase flow phenomena and for model validation purposes. This is reflected also in this issue, where only one article is purely experimental, while three of them deal with theoretical modelling and the remaining three with numerical simulations. The experimental investigation of the critical heat flux (CHF) phenomena by means of photographic study is presented in the paper of J. Park et al. They have used a high-speed camera system to observe the transient boiling characteristics on a thin horizontal cylinder submerged in a pool of water or highly wetting liquid. Experiments show that the initial boiling process is strongly affected by the properties and wettability of the liquid. The authors have stressed the importance of the local scale observation leading to better understanding of the transient CHF phenomena. In the article of G. Espinosa-Paredes et al. a theoretical work concerning the derivation of transport equations for two-phase flow is presented. The author proposes a novel approach based on derivation of nonlocal volume averaged equations which contain new terms related to nonlocal transport effects. These non-local terms act as coupling elements between the phenomena

  16. Latent heat transport and microlayer evaporation in nucleate boiling

    International Nuclear Information System (INIS)

    Jawurek, H.H.

    1977-08-01

    Part 1 of this work provides a broad overview and, where possible, a quantitative assessment of the complex physical processes which together constitute the mechanism of nucleate boiling heat transfer. It is shown that under a wide range of conditions the primary surface-to-liquid heat flows within an area of bubble influence are so redistributed as to manifest themselves predominantly as latent heat transport, that is, as vaporisation into attached bubbles. Part 2 deals in greater detail with one of the component processes of latent heat transport, namely microlayer evaporation. A literature review reveals the need for synchronised records of microlayer geometry versus time and of normal bubble growth and departure. An apparatus developed to provide such records is described. High-speed cine interference photography from beneath and through a transparent heating surface provided details of microlayer geometry and an image reflection system synchronised these records with the bubble profile views. Results are given for methanol and ethanol boiling at sub-atmospheric pressures and at various heat fluxes and bulk subcoolings. In all cases it is found that microlayers were of sub-micron thickness, that microlayer thinning was restricted to the inner layer edge (with the thickness elsewhere remaining constant or increasing with time) and that the contribution of this visible evaporation to the total vapour flow into bubbles was negligible. The observation of thickening towards the outer microlayer edge, however, demonstrates that a liquid replenishment flow occurred simultaneously with the evaporation process

  17. Changes in ocean vertical heat transport with global warming

    Science.gov (United States)

    Zika, Jan D.; Laliberté, Frédéric; Mudryk, Lawrence R.; Sijp, Willem P.; Nurser, A. J. G.

    2015-06-01

    Heat transport between the surface and deep ocean strongly influences transient climate change. Mechanisms setting this transport are investigated using coupled climate models and by projecting ocean circulation into the temperature-depth diagram. In this diagram, a "cold cell" cools the deep ocean through the downwelling of Antarctic waters and upwelling of warmer waters and is balanced by warming due to a "warm cell," coincident with the interhemispheric overturning and previously linked to wind and haline forcing. With anthropogenic warming, the cold cell collapses while the warm cell continues to warm the deep ocean. Simulations with increasingly strong warm cells, set by their mean Southern Hemisphere winds, exhibit increasing deep-ocean warming in response to the same anthropogenic forcing. It is argued that the partition between components of the circulation which cool and warm the deep ocean in the preindustrial climate is a key determinant of ocean vertical heat transport with global warming.

  18. The new context for transport of radioactive and nuclear material

    International Nuclear Information System (INIS)

    Anne, C.; Galtier, J.

    2002-01-01

    The transportation of radioactive and nuclear materials involves all modes of transportation with a predominance for road and for air. It is but a minute fraction dangerous good transportation. Around 10 millions of radioactive packages are shipped annually all over the world of which ninety percent total corresponds to shipments of radioisotopes. In spite of the small volume transported, experience, evolution of transport means and technologies, the trend to constantly improve security and safety and public acceptance have modified the transport environment. During the last few years, new evolutions have applied to the transport of radioactive and nuclear materials in various fields and especially: - Safety - Security - Logistics means - Public acceptance - Quality Assurance. We propose to examine the evolution of these different fields and their impact on transportation methods and means. (authors)

  19. Application possibilities for nuclear heating plants in the energy system of the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Kohler, T.

    1991-01-01

    The field of application for nuclear heating plants is the so-called low-temperature heating market. It includes the energy demand for space heating, hot water an low-temperature process heat. The analysis of technical potentials for heating reactors considers two different levels. The structure of the district heating system determines the technical potential in the now existing energy system, it amounts to a total power of 9,8 to 14,3 GW th of heating reactors. For a possible extended use of heating reactors in future which goes beyond the existing district heating system the technical circumstances and the local distribution of the low-temperature heating market define the technical potential which ranges from 126 to 160 GW th on todays basis. The chance of implementing nuclear heating plants is strongly influenced by the economy of their heat generation. The economic situation of heat generation with heating reactors is estimated in comparison to current fossil district heating production systems. In the low-temperature heating market the heat supply by nuclear fed district heating systems is compared to the heat production in houses. Considering the assumptions the analysis indicates that nuclear heating plants can compete with existing fossil heat sources. The analysis shows that heating reactors are an interesting and powerful option for the supply of the district heating market in future. The underlying economic assumptions would allow the use of nuclear heating plants and it seems that they could contribute to reduce the environmental stress. (orig.) [de

  20. Perturbative Heat Transport Experiments on TJ-II

    International Nuclear Information System (INIS)

    Eguilor, S.; Castejon, F.; Luna, E. de la; Cappa, A.; Likin, K.; Fernandez, A.; Tj-II, T.

    2002-01-01

    Heat wave experiments are performed on TJ-II stellarator plasmas to estimate both heat diffusivity and power deposition profiles. High frequency ECRH modulation experiments are used to obtain the power deposition profiles, which is observed to be wider and duller than estimated by tracing techniques. The causes of this difference are discussed in the paper. Fourier analysis techniques are used to estimate the heat diffusivity in low frequency ECRH modulation experiments. This include the power deposition profile as a new ingredient. ECHR switch on/off experiments are exploited to obtain power deposition and heat diffusivities profile. Those quantities are compared with the obtained by modulation experiments and transport analysis, showing a good agreement. (Author) 18 refs

  1. Core heat transport in the MAST Spherical Tokamak

    International Nuclear Information System (INIS)

    Field, A.R.; Akers, R.J.; Brickley, C.; Carolan, P.G.; Challis, C.; Conway, N.J.; Cunningham, G.; Meyer, H.; Patel, A.; Roach, C.; Valovie, M.; Applegate, D.J.; Cowley, S.C.; Joiner, N.; Walsh, M.J.

    2005-01-01

    High-β spherical tokamak (ST) plasmas have intrinsic properties which favour the suppression of anomalous transport. Transport has been studied in NBI heated plasmas in the MAST ST device, where it is found that ion thermal transport is typically close to the neo-classical level. Calculations of the ITG microstability with the GS2 gyro-kinetic code suggest that this form of turbulence may be suppressed by the high ExB shearing rates in these plasmas. Electron transport is somewhat higher and cannot be explained from mixing length estimates of ETG turbulence. This is perhaps due instead either to micro-tearing modes in the core plasma or extended radial structures in the saturated turbulence. Micro-stability is also favoured by low magnetic shear and this has been used to produce high-performance L- and H-mode plasmas with improved core confinement as well as plasmas exhibiting ITBs in both the ion and electron channels. Broad electron ITBs have been produced with counter-NBI heating in which anomalous electron transport apparently has been reduced by the very high ExB shearing rates prevailing in these plasmas. Such studies also contribute towards testing the transport and ITB physics basis for the ITER device. (author)

  2. Promising design options for the encapsulated nuclear heat source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

    2001-07-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  3. Promising design options for the encapsulated nuclear heat source reactor

    International Nuclear Information System (INIS)

    Conway, L.; Carelli, M.D.; Dzodzo, M.; Hossain, Q.; Brown, N.W.; Wade, D.C.; Sienick, J.J.; Greenspan, E.; Kastenberg, W.E.; Saphier, D.

    2001-01-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  4. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

  5. Unsteady heat exchange at the dry spent nuclear fuel storage

    OpenAIRE

    Svitlana Alyokhina; Andrii Kostikov

    2017-01-01

    Unsteady thermal processes in storage containers with spent nuclear fuel were modeled. The daily fluctuations of outer ambient temperatures were taken into account. The modeling approach, which is based on the solving of conjugate and inverse heat transfer problems, was verified by comparison of measured and calculated temperatures in outer channels. The time delays in the reaching of maximal temperatures for each spent fuel assembly were calculated. Results of numerical investigations show t...

  6. Nuclear heating reactor, an advanced and passive reactor

    International Nuclear Information System (INIS)

    Wang Dazhong; Zheng Wenxiang

    1994-01-01

    The nuclear heating reactor (NHR) is designed with a number of the advanced and innovative features, including integrated arrangement, natural circulation, self-pressurized performance, dual vessel structure, hydraulic control rod drive and passive safety systems. Being an advanced and passive reactor, the NHR can serve as a clean, safe and economic energy source. This paper describes the development status, main design and safety features of the NHR. 3 refs., 2 tabs., 5 figs

  7. Potential industrial market for process heat from nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, R.W.

    1976-07-01

    A specific segment of industrial process heat use has been examined in detail to identify individual plant locations throughout the United states where nuclear generated steam may be a viable alternative. Five major industries have been studied: paper, chemicals, petroleum, rubber, and primary metals. For these industries, representing 75 percent of the total industrial steam consumption, the individual plant locations within the U.S. using steam in large quantities have been located and characterized as to fuel requirements

  8. Next Generation Nuclear Plant Intermediate Heat Exchanger Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C to 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium cooled, prismatic or pebble-bed reactor, and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Intermediate Heat Exchanger (IHX).This component will be operated in flowing, impure helium on the primary and secondary side at temperatures up to 950°C. There are major high temperature design, materials availability, and fabrication issues that need to be addressed. The prospective materials are Alloys 617, 230, 800H and X, with Alloy 617 being the leading candidate for the use at 950°C. The material delivery schedule for these materials does not pose a problem for a 2018 start up as the vendors can quote reasonable delivery times at the moment. The product forms and amount needed must be finalized as soon as possible. An

  9. The new context for transport of radioactive nuclear material

    International Nuclear Information System (INIS)

    Anne, Catherine; Galtier, Jerome

    2001-01-01

    The transportation of radioactive and nuclear materials, involves all modes of transportation (road, air, sea, rail) with predominance for road and for air (air for radioisotopes). In this paper we examine the impact of new evolutions in the fields of safety, security, logistics means, public acceptance and quality assurance

  10. Heat transport modelling in EXTRAP T2R

    Science.gov (United States)

    Frassinetti, L.; Brunsell, P. R.; Cecconello, M.; Drake, J. R.

    2009-02-01

    A model to estimate the heat transport in the EXTRAP T2R reversed field pinch (RFP) is described. The model, based on experimental and theoretical results, divides the RFP electron heat diffusivity χe into three regions, one in the plasma core, where χe is assumed to be determined by the tearing modes, one located around the reversal radius, where χe is assumed not dependent on the magnetic fluctuations and one in the extreme edge, where high χe is assumed. The absolute values of the core and of the reversal χe are determined by simulating the electron temperature and the soft x-ray and by comparing the simulated signals with the experimental ones. The model is used to estimate the heat diffusivity and the energy confinement time during the flat top of standard plasmas, of deep F plasmas and of plasmas obtained with the intelligent shell.

  11. Heating control device for nuclear reactor cooling systems

    International Nuclear Information System (INIS)

    Kishigawa, Osamu.

    1981-01-01

    Purpose: To obtain a heating control device capable of surely preventing deformations and fluctures due to thermal stresses in equipments such as tanks and pipeways upon starting of reactor operation. Constitution: A heating control device for nuclear reactor cooling systems using metal coolants comprises a plurality of heaters disposed at each of the sections in the cooling systems for heating the coolants, temperature detectors for the detection of temperature at each of the sections in the cooling systems to be heated by the heaters, a circuit for judging the range of filling process of the metal coolants in the cooling systems, a coefficient change circuit for heating control for changing and setting the coefficient for the heating control by the heaters based on the information for the range judged by the judging circuit, and a circuit for controlling the input to each of the heaters based on the output signals for the coefficient change circuit and the signals from the temperature detectors. (Seki, T.)

  12. Waste heat from nuclear power stations-challenge to agriculture

    International Nuclear Information System (INIS)

    Horacek, P.

    1981-01-01

    The possibilities are considered of using low-grade waste heat from nuclear power plants as a source of energy. Potential areas are explored of the waste heat applications and three main fields are defined where the use of waste heat would not only be beneficial but also feasible, namely heating green-houses, warming agricultural soils and warming water reservoirs used for fishery. Examples are given of experimental use of waste heat in the USSR, the USA, France, the GDR, the FRG, Hungary, Poland and elsewhere for agricultural purposes. The results are reported of cost-benefit analyses showing that although this type of waste heat application would be most feasible, economic obstacles have so far hindered its spread. Before it can be utilized on an industrial scale, the problems should be resolved of high cost, of harmful consequences of increasing the temperature of water and soils, such as reduced resistance of plants and fish to disease or increased activity of insects, and of the environmental impact. (L.O.)

  13. Coupled analysis of flow and heat around a high-level nuclear waste repository

    International Nuclear Information System (INIS)

    Utsugida, Y.

    1985-01-01

    This paper presents the formulation of coupled flow and heat models by F.E.M. and the results of two kinds of analysis around a high-level nuclear waste repository. A one-dimensional element is used in modeling fluid flow in fractured media, but two-dimensional elements are used in modeling fluid flow and heat transport in porous media. A 2-order Runge-Kutta scheme is employed to carry out the time integration, and the coefficients of the capacity matrix are lumped at the nodes. One of the analyses is in the near field which regards disposal tunnel scale. The effects of the fluid flow are investigated on the temperature distribution in the vicinity of a disposal tunnel. The other is in the far field which regards the regional scale of the entire repository. The effects of radio-genic heat being produced by the nuclear wastes are investigated on the ground water flow in the cases of major fractured rock and no major fractured rock. In conclusion, there is only a slight influence from fluid flow in the near field and from heat in the far field in the case of no major fractured rock. On the other hand, the influence of radio-genic heat in the far field in the case of major fractured rock can not be ignored

  14. Quantum Thermodynamics in Strong Coupling: Heat Transport and Refrigeration

    Directory of Open Access Journals (Sweden)

    Gil Katz

    2016-05-01

    Full Text Available The performance characteristics of a heat rectifier and a heat pump are studied in a non-Markovian framework. The device is constructed from a molecule connected to a hot and cold reservoir. The heat baths are modelled using the stochastic surrogate Hamiltonian method. The molecule is modelled by an asymmetric double-well potential. Each well is semi-locally connected to a heat bath composed of spins. The dynamics are driven by a combined system–bath Hamiltonian. The temperature of the baths is regulated by a secondary spin bath composed of identical spins in thermal equilibrium. A random swap operation exchange spins between the primary and secondary baths. The combined system is studied in various system–bath coupling strengths. In all cases, the average heat current always flows from the hot towards the cold bath in accordance with the second law of thermodynamics. The asymmetry of the double well generates a rectifying effect, meaning that when the left and right baths are exchanged the heat current follows the hot-to-cold direction. The heat current is larger when the high frequency is coupled to the hot bath. Adding an external driving field can reverse the transport direction. Such a refrigeration effect is modelled by a periodic driving field in resonance with the frequency difference of the two potential wells. A minimal driving amplitude is required to overcome the heat leak effect. In the strong driving regime the cooling power is non-monotonic with the system–bath coupling.

  15. Nuclear reactor fuel element having improved heat transfer

    Science.gov (United States)

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  16. JOYO MK-III heat transport system renovation operation. Primary heat transport mechanical system (IHXs (intermediate heat exchangers))

    International Nuclear Information System (INIS)

    Ohshima, Jun; Ashida, Takashi; Isozaki, Kazunori; Sumino, Kouzou; Yamaguchi, Akira; Sakaba, Hideo; Tomita, Naoki; Ozawa, Kenji

    2004-04-01

    The MK-III project to improve the irradiation capability of the experimental fast reactor JOYO have been carried out since 1987. The increase of fast neutron flux and the enlargement of irradiation field increase the reactor thermal power from 100 MWt to 140 MWt. To accommodate the increased thermal power, the IHXs and the IHX connecting piping were replaced. The IHXs were replaced with securing cooling system boundary in high dose rate surroundings and very limited operation space of the radiation controlled area in the containment vessel. Primary sodium contains radioactive 22 Na, 24 Na and radioactive CPs such as 60 Co and 54 Mn, and this sodium adhered to the inner surface of IHXs and pipe. Therefore, the renovation procedure and method were carefully examined based on the JOYO operation and maintenance experiences and research and development results on the sodium handling technique. The major results obtained in the primary heat transport mechanical system (IHXs) renovation operation were shown as follows; (1) The mock up tests to optimize the operating methods, to check the operability and for workers training were useful for reduction of radiation exposure by shortening the operation time in high dose rate surrounding. (2) The effectiveness of seal bag for prevention of impurity ingress to the sodium system and contamination during sodium boundary opening (cutting pipes, sodium removal and welding pipes) was confirmed. (3) The pipes were cut without foreign object such as cutting piece and tool ingress by careful examination of cutting procedure and methods such as bite, roller cutter. (4) The temporary closing equipment such as seal cap and seal plug were effectively worked to seal the cooling system boundary between cutting and welding pipes. (5) Sodium adhered on the inner surfaces of pipe was effectively and safely removed by a mechanical scraper or drill and a cloth moistened by a mixture of alcohol and water. (6) Control of low gas pressure difference

  17. Thermoacoustic sensor for nuclear fuel temperaturemonitoring and heat transfer enhancement

    Energy Technology Data Exchange (ETDEWEB)

    James A. Smith; Dale K. Kotter; Randall A. Alli; Steven L. Garrett

    2013-05-01

    A new acoustical sensing system for the nuclear power industry has been developed at The Pennsylvania State University in collaboration with Idaho National Laboratories. This sensor uses the high temperatures of nuclear fuel to convert a nuclear fuel rod into a standing-wave thermoacoustic engine. When a standing wave is generated, the sound wave within the fuel rod will be propagated, by acoustic radiation, through the cooling fluid within the reactor or spent fuel pool and can be monitored a remote location external to the reactor. The frequency of the sound can be correlated to an effective temperature of either the fuel or the surrounding coolant. We will present results for a thermoacoustic resonator built into a Nitonic-60 (stainless steel) fuel rod that requires only one passive component and no heat exchangers.

  18. Convective heat transport in Viscoplastic material due to localized heating: An Experimental approach

    Science.gov (United States)

    Jhunjhunwala, Kishan; Shahiruddin; Hassan, M. A.

    2018-02-01

    Viscoplastic materials are found extensively both in natural and manmade form. In this work experimental investigation of natural convection in viscoplastic fluid with partially heated bottom wall and continuous cooling of top wall in a square cross section enclosure has been carried out. Carbopol Ultrez 20 gel of various concentrations has been used as sample viscoplastic fluid. Conduction and convection phase of heat transport are identified. The results are presented in terms of temperature distribution across the fluid for different gel concentrations and heat input. The average Nusselt numbers are also discussed for different conditions. Onset of convection is delayed and convection strength is weakened with increase in test fluid yield stress. Steady state temperature difference between hot and cold wall shows linear behaviour with heat input for conduction regime and non-linear behaviour in convection regime. Fluid temperature in enclosure shows sharp gradient closure to thermally active walls.

  19. The Yeast Nuclear Pore Complex and Transport Through It

    Science.gov (United States)

    Aitchison, John D.; Rout, Michael P.

    2012-01-01

    Exchange of macromolecules between the nucleus and cytoplasm is a key regulatory event in the expression of a cell’s genome. This exchange requires a dedicated transport system: (1) nuclear pore complexes (NPCs), embedded in the nuclear envelope and composed of proteins termed nucleoporins (or “Nups”), and (2) nuclear transport factors that recognize the cargoes to be transported and ferry them across the NPCs. This transport is regulated at multiple levels, and the NPC itself also plays a key regulatory role in gene expression by influencing nuclear architecture and acting as a point of control for various nuclear processes. Here we summarize how the yeast Saccharomyces has been used extensively as a model system to understand the fundamental and highly conserved features of this transport system, revealing the structure and function of the NPC; the NPC’s role in the regulation of gene expression; and the interactions of transport factors with their cargoes, regulatory factors, and specific nucleoporins. PMID:22419078

  20. Demonstrating Hybrid Heat Transport and Energy Conversion System Performance Characterization Using Intelligent Control Systems

    International Nuclear Information System (INIS)

    Ostrum, Lee; Manic, Milos

    2017-01-01

    The debate continues on the magnitude and validity of climate change caused by human activities. However, there is no debate about the need to make buildings, modes of transportation, factories, and homes as energy efficient as possible. Given that climate change could occur with the wasteful use of fossil fuel and the fact that fossil energy costs could and will swing wildly, it is imperative that every effort be made to utilize energy sources to their fullest. Hybrid energy systems (HES) are two or more separate energy producers used together to produce energy commodities. The HES this report focuses on is the use of nuclear reactor waste heat as a source of further energy utilization. Nuclear reactors use a fluid to cool the core and produce the steam needed for the production of electricity. Traditionally this steam, or coolant, is used to convert the energy then cooled elsewhere. The heat is released into the environment without being used further. By adding technologies to nuclear reactors to use the wasted heat, a system can be developed to make more than just electricity and allow for loading following capabilities.

  1. Demonstrating Hybrid Heat Transport and Energy Conversion System Performance Characterization Using Intelligent Control Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ostrum, Lee [Univ. of Idaho and Idaho Falls Center, Idaho Falls, ID (United States); Manic, Milos [Virginia Commonwealth Univ., Richmond, VA (United States)

    2017-09-28

    The debate continues on the magnitude and validity of climate change caused by human activities. However, there is no debate about the need to make buildings, modes of transportation, factories, and homes as energy efficient as possible. Given that climate change could occur with the wasteful use of fossil fuel and the fact that fossil energy costs could and will swing wildly, it is imperative that every effort be made to utilize energy sources to their fullest. Hybrid energy systems (HES) are two or more separate energy producers used together to produce energy commodities. The HES this report focuses on is the use of nuclear reactor waste heat as a source of further energy utilization. Nuclear reactors use a fluid to cool the core and produce the steam needed for the production of electricity. Traditionally this steam, or coolant, is used to convert the energy then cooled elsewhere. The heat is released into the environment without being used further. By adding technologies to nuclear reactors to use the wasted heat, a system can be developed to make more than just electricity and allow for loading following capabilities.

  2. Transport lattice models of heat transport in skin with spatially heterogeneous, temperature-dependent perfusion

    Directory of Open Access Journals (Sweden)

    Martin Gregory T

    2004-11-01

    Full Text Available Abstract Background Investigation of bioheat transfer problems requires the evaluation of temporal and spatial distributions of temperature. This class of problems has been traditionally addressed using the Pennes bioheat equation. Transport of heat by conduction, and by temperature-dependent, spatially heterogeneous blood perfusion is modeled here using a transport lattice approach. Methods We represent heat transport processes by using a lattice that represents the Pennes bioheat equation in perfused tissues, and diffusion in nonperfused regions. The three layer skin model has a nonperfused viable epidermis, and deeper regions of dermis and subcutaneous tissue with perfusion that is constant or temperature-dependent. Two cases are considered: (1 surface contact heating and (2 spatially distributed heating. The model is relevant to the prediction of the transient and steady state temperature rise for different methods of power deposition within the skin. Accumulated thermal damage is estimated by using an Arrhenius type rate equation at locations where viable tissue temperature exceeds 42°C. Prediction of spatial temperature distributions is also illustrated with a two-dimensional model of skin created from a histological image. Results The transport lattice approach was validated by comparison with an analytical solution for a slab with homogeneous thermal properties and spatially distributed uniform sink held at constant temperatures at the ends. For typical transcutaneous blood gas sensing conditions the estimated damage is small, even with prolonged skin contact to a 45°C surface. Spatial heterogeneity in skin thermal properties leads to a non-uniform temperature distribution during a 10 GHz electromagnetic field exposure. A realistic two-dimensional model of the skin shows that tissue heterogeneity does not lead to a significant local temperature increase when heated by a hot wire tip. Conclusions The heat transport system model of the

  3. Transport of nuclear material (Part II)

    International Nuclear Information System (INIS)

    Staake, Theo; Schmidt, Thomas

    1983-01-01

    Providing a complete back-end service for MTR reactors is one of the fundamental and traditional tasks of TRANSNUKLEAR GmbH (TN). TN's services in this field cover everything from supplying the ideal transport cask, providing technical assistance during the loading operation, obtaining the necessary package approval and transport licenses, providing the required insurance cover, carrying out the transport, right thru to settling the reprocessing contract. Up until 1976, TN carried out transports of MTR fuel elements to the European reprocessing plants at Mol in Belgium and Marcoule in France. In all, some 1000 fuel elements were transported in this p e ri od. However, following the decision by these plants not to reprocess these elements anymore, subsequent transports had to be made to the US-DOE reprocessing plants. TN pooled together the interests of all her MTR customers and signed a reprocessing contract with the US-DOE, which ensured a complete back-end service for these reactors well into the future. In close cooperation with our associated company, Transnuclear Inc. in New York and Washington, a new transport concept was developed, which proved itself to be both economic and reliable. Up to now, a total of about 2050 MTR fuel elements have been transported by TN-Germany to the USA in 65 separate shipments. The total number of shipments performed by the TN group is 165 shipments. All shipments were carried out routinely without any incident. In March this year, the US-DOE made use of a clause in the contract, in which 90 days' notice was given of a change in reprocessing plant. Whereas previously all elements had been taken to the Savannah River Plant (SRP) in South Carolina, in future all elements have to go to the Idaho Chemical Processing Plant (ICPP) near Idaho Falls. This change presented TN with the not inconsiderable problem of finding a suitable transport route. Due to the large number of influencing factors, the TN-group carried out a special

  4. Transportation capabilities study of DOE-owned spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1994-10-01

    This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

  5. Salt disposal of heat-generating nuclear waste

    International Nuclear Information System (INIS)

    Leigh, Christi D.; Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

  6. Salt disposal of heat-generating nuclear waste.

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

  7. Regional waste treatment facilities with underground monolith disposal for all low-heat-generating nuclear wastes

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1982-01-01

    An alternative system for treatment and disposal of all ''low-heat-generating'' nuclear wastes from all sources is proposed. The system, Regional Waste Treatment Facilities with Underground Monolith Disposal (RWTF/UMD), integrates waste treatment and disposal operations into single facilities at regional sites. Untreated and/or pretreated wastes are transported from generation sites such as reactors, hospitals, and industries to regional facilities in bulk containers. Liquid wastes are also transported in bulk after being gelled for transport. The untreated and pretreated wastes are processed by incineration, crushing, and other processes at the RWTF. The processed wastes are mixed with cement. The wet concrete mixture is poured into large low-cost, manmade caverns or deep trenches. Monolith dimensions are from 15 to 25 m wide, and 20 to 60 m high and as long as required. This alternative waste system may provide higher safety margins in waste disposal at lower costs

  8. Probabilistic Risk Assessment on Maritime Spent Nuclear Fuel Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Christian, Robby; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    Spent nuclear fuel (SNF) management has been an indispensable issue in South Korea. Before a long term SNF solution is implemented, there exists the need to distribute the spent fuel pool storage loads. Transportation of SNF assemblies from populated pools to vacant ones may preferably be done through the maritime mode since all nuclear power plants in South Korea are located at coastal sites. To determine its feasibility, it is necessary to assess risks of the maritime SNF transportation. This work proposes a methodology to assess the risk arising from ship collisions during the transportation of SNF by sea. Its scope is limited to the damage probability of SNF packages given a collision event. The effect of transport parameters' variation to the package damage probability was investigated to obtain insights into possible ways to minimize risks. A reference vessel and transport cask are given in a case study to illustrate the methodology's application.

  9. Heterogeneous nuclear ribonucleoprotein K inhibits heat shock-induced transcriptional activity of heat shock factor 1.

    Science.gov (United States)

    Kim, Hee-Jung; Lee, Jae-Jin; Cho, Jin-Hwan; Jeong, Jaeho; Park, A Young; Kang, Wonmo; Lee, Kong-Joo

    2017-08-04

    When cells are exposed to heat shock and various other stresses, heat shock factor 1 (HSF1) is activated, and the heat shock response (HSR) is elicited. To better understand the molecular regulation of the HSR, we used 2D-PAGE-based proteome analysis to screen for heat shock-induced post-translationally modified cellular proteins. Our analysis revealed that two protein spots typically present on 2D-PAGE gels and containing heterogeneous nuclear ribonucleoprotein K (hnRNP K) with trioxidized Cys 132 disappeared after the heat shock treatment and reappeared during recovery, but the total amount of hnRNP K protein remained unchanged. We next tested whether hnRNP K plays a role in HSR by regulating HSF1 and found that hnRNP K inhibits HSF1 activity, resulting in reduced expression of hsp70 and hsp27 mRNAs. hnRNP K also reduced binding affinity of HSF1 to the heat shock element by directly interacting with HSF1 but did not affect HSF1 phosphorylation-dependent activation or nuclear localization. hnRNP K lost its ability to induce these effects when its Cys 132 was substituted with Ser, Asp, or Glu. These findings suggest that hnRNP K inhibits transcriptional activity of HSF1 by inhibiting its binding to heat shock element and that the oxidation status of Cys 132 in hnRNP K is critical for this inhibition. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

  10. Three dimensional heat transport modeling in Vossoroca reservoir

    Science.gov (United States)

    Arcie Polli, Bruna; Yoshioka Bernardo, Julio Werner; Hilgert, Stephan; Bleninger, Tobias

    2017-04-01

    Freshwater reservoirs are used for many purposes as hydropower generation, water supply and irrigation. In Brazil, according to the National Energy Balance of 2013, hydropower energy corresponds to 70.1% of the Brazilian demand. Superficial waters (which include rivers, lakes and reservoirs) are the most used source for drinking water supply - 56% of the municipalities use superficial waters as a source of water. The last two years have shown that the Brazilian water and electricity supply is highly vulnerable and that improved management is urgently needed. The construction of reservoirs affects physical, chemical and biological characteristics of the water body, e.g. stratification, temperature, residence time and turbulence reduction. Some water quality issues related to reservoirs are eutrophication, greenhouse gas emission to the atmosphere and dissolved oxygen depletion in the hypolimnion. The understanding of the physical processes in the water body is fundamental to reservoir management. Lakes and reservoirs may present a seasonal behavior and stratify due to hydrological and meteorological conditions, and especially its vertical distribution may be related to water quality. Stratification can control heat and dissolved substances transport. It has been also reported the importance of horizontal temperature gradients, e.g. inflows and its density and processes of mass transfer from shallow to deeper regions of the reservoir, that also may impact water quality. Three dimensional modeling of the heat transport in lakes and reservoirs is an important tool to the understanding and management of these systems. It is possible to estimate periods of large vertical temperature gradients, inhibiting vertical transport and horizontal gradients, which could be responsible for horizontal transport of heat and substances (e.g. differential cooling or inflows). Vossoroca reservoir was constructed in 1949 by the impoundment of São João River and is located near to

  11. HTTR demonstration test plan for industrial utilization of nuclear heat

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Yan, Xing L.; Kubo, Shinji; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

    2014-09-01

    Japan Atomic Energy Agency has been conducting research and development with a central focus on the utilization of High Temperature engineering Test Reactor (HTTR), the first High Temperature Gas-cooled Reactor (HTGR) in Japan, towards the realization of industrial use of nuclear heat. Several studies have made on the integration of the HTTR with thermochemical iodine-sulfur process and steam methane reforming hydrogen production plant (H 2 plant) as well as helium gas turbine power conversion system. In addition, safety standards for coupling a H 2 plant to a nuclear facility has been investigated. Based on the past design information, the present study identified test items to be validated in the HTTR demonstration test to accomplish a formulation of safety requirement and design consideration for coupling a H 2 plant to a nuclear facility as well as confirmation of overall performance of helium gas turbine system. In addition, plant concepts for the heat utilization system to be connected with the HTTR are investigated. (author)

  12. Colloid transport code-nuclear user's manual

    International Nuclear Information System (INIS)

    Jain, R.

    1992-01-01

    This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems

  13. Pulsating Heat Pipes, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — An advanced heat transport technology is presented that can enable space nuclear power systems to transfer reactor heat, convert heat into electricity, reject waste...

  14. Experimental determination of soil heat storage for the simulation of heat transport in a coastal wetland

    Science.gov (United States)

    Swain, Michael; Swain, Matthew; Lohmann, Melinda; Swain, Eric

    2012-01-01

    Two physical experiments were developed to better define the thermal interaction of wetland water and the underlying soil layer. This information is important to numerical models of flow and heat transport that have been developed to support biological studies in the South Florida coastal wetland areas. The experimental apparatus consists of two 1.32. m diameter by 0.99. m tall, trailer-mounted, well-insulated tanks filled with soil and water. A peat-sand-soil mixture was used to represent the wetland soil, and artificial plants were used as a surrogate for emergent wetland vegetation based on size and density observed in the field. The tanks are instrumented with thermocouples to measure vertical and horizontal temperature variations and were placed in an outdoor environment subject to solar radiation, wind, and other factors affecting the heat transfer. Instruments also measure solar radiation, relative humidity, and wind speed.Tests indicate that heat transfer through the sides and bottoms of the tanks is negligible, so the experiments represent vertical heat transfer effects only. The temperature fluctuations measured in the vertical profile through the soil and water are used to calibrate a one-dimensional heat-transport model. The model was used to calculate the thermal conductivity of the soil. Additionally, the model was used to calculate the total heat stored in the soil. This information was then used in a lumped parameter model to calculate an effective depth of soil which provides the appropriate heat storage to be combined with the heat storage in the water column. An effective depth, in the model, of 5.1. cm of wetland soil represents the heat storage needed to match the data taken in the tank containing 55.9. cm of peat/sand/soil mix. The artificial low-density laboratory sawgrass reduced the solar energy absorbed by the 35.6. cm of water and 55.9. cm of soil at midday by less than 5%. The maximum heat transfer into the underlying peat-sand-soil mix

  15. Cascade: a review of heat transport and plant design issues

    International Nuclear Information System (INIS)

    Murray, K.A.; McDowell, M.W.

    1984-01-01

    A conceptual heat transfer loop for Cascade, a centrifugal-action solid-breeder reaction chamber, has been investigated and results are presented. The Cascade concept, a double-cone-shaped reaction chamber, rotates along its horizontal axis. Solid Li 2 O or other lithium-ceramic granules are injected tangentially through each end of the chamber. The granules cascade axially from the smaller radii at the ends to the larger radius at the center, where they are ejected into a stationary granule catcher. Heat and tritium are then removed from the granules and the granules are reinjected into the chamber. A 50% dense Li 2 O granule throughput of 2.8 m 3 /s is transferred from the reaction chamber to the steam generators via continuous bucket elevators. The granules then fall by gravity through 4 vertical steam generators. The entire transport system is maintained at the same vacuum conditions present inside the reaction chamber

  16. Energetics of Transport through the Nuclear Pore Complex.

    Directory of Open Access Journals (Sweden)

    Ali Ghavami

    Full Text Available Molecular transport across the nuclear envelope in eukaryotic cells is solely controlled by the nuclear pore complex (NPC. The NPC provides two types of nucleocytoplasmic transport: passive diffusion of small molecules and active chaperon-mediated translocation of large molecules. It has been shown that the interaction between intrinsically disordered proteins that line the central channel of the NPC and the transporting cargoes is the determining factor, but the exact mechanism of transport is yet unknown. Here, we use coarse-grained molecular dynamics simulations to quantify the energy barrier that has to be overcome for molecules to pass through the NPC. We focus on two aspects of transport. First, the passive transport of model cargo molecules with different sizes is studied and the size selectivity feature of the NPC is investigated. Our results show that the transport probability of cargoes is significantly reduced when they are larger than ∼5 nm in diameter. Secondly, we show that incorporating hydrophobic binding spots on the surface of the cargo effectively decreases the energy barrier of the pore. Finally, a simple transport model is proposed which characterizes the energy barrier of the NPC as a function of diameter and hydrophobicity of the transporting particles.

  17. Electron heat transport analysis of low-collisionality plasmas in the neoclassical-transport-optimized configuration of LHD

    International Nuclear Information System (INIS)

    Murakami, Sadayoshi; Yamada, Hiroshi; Wakasa, Arimitsu

    2002-01-01

    Electron heat transport in low-collisionality LHD plasma is investigated in order to study the neoclassical transport optimization effect on thermal plasma transport with an optimization level typical of so-called ''advanced stellarators''. In the central region, a higher electron temperature is obtained in the optimized configuration, and transport analysis suggests the considerable effect of neoclassical transport on the electron heat transport assuming the ion-root level of radial electric field. The obtained experimental results support future reactor design in which the neoclassical and/or anomalous transports are reduced by magnetic field optimization in a non-axisymmetric configuration. (author)

  18. Coupling of high temperature nuclear reactor with chemical plant by means of steam loop with heat pump

    Directory of Open Access Journals (Sweden)

    Kopeć Mariusz

    2017-01-01

    Full Text Available High temperature nuclear reactors (HTR can be used as an excellent, emission-free source of technological heat for various industrial applications. Their outlet helium temperature (700°-900°C allows not only for heat supply to all processes below 600°C (referred to as “steam class”, but also enables development of clean nuclear-assisted hydrogen production or coal liquefaction technologies with required temperatures up to 900°C (referred to as “chemical class”. This paper presents the results of analyses done for various configurations of the steam transport loop coupled with the high-temperature heat pump designed for “chemical class” applications. The advantages and disadvantages as well as the key issues are discussed in comparison with alternative solutions, trying to answer the question whether the system with the steam loop and the hightemperature heat pump is viable and economically justified.

  19. Heat transport inventory monitoring for CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Hussein, E.; Luxat, J.C.

    1984-01-01

    A computer-based D 2 O coolant inventory monitoring system proposed for implementation on the digital computer controllers at Ontario Hydro's CANDU generating units is discussed. By monitoring process parameters and utilizing probabilistically-based decision algorithms, timely indication of any significant loss of D 2 O inventory will be provided to the operator. The monitoring is performed in a co-ordinated manner such that D 2 O losses from either the heat transport system or the inventory control system can be detected. (orig.)

  20. Aging management guideline for commercial nuclear power plants - heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Booker, S.; Lehnert, D.; Daavettila, N.; Palop, E.

    1994-06-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in commercial nuclear power plant heat exchangers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  1. Synthesis of hydrocarbons using coal and nuclear process heat

    International Nuclear Information System (INIS)

    Eickhoff, H.G.; Kugeler, K.

    1975-01-01

    An analysis of the global petroleum resources and demand shows that the amount of mineral oil products is sufficient to meet the requirements of the next decades. The geographical resources, however, could lead to problems of distribution and foreign exchange. The production of hydrocarbons with coal as basis using high temperature nuclear process heat has advantages compared to the conventional techniques. Next to the conservation of reserve fossil primary energy carriers there are advantages as regards prices, which at high coal costs are especially pronounced. (orig.) [de

  2. Aging management guideline for commercial nuclear power plants - heat exchangers

    International Nuclear Information System (INIS)

    Booker, S.; Lehnert, D.; Daavettila, N.; Palop, E.

    1994-06-01

    This Aging Management Guideline (AMG) describes recommended methods for effective detection and mitigation of age-related degradation mechanisms in commercial nuclear power plant heat exchangers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein

  3. Yucca Mountain nuclear waste repository prompts heated congressional hearing

    Science.gov (United States)

    Showstack, Randy

    2011-11-01

    Although the final report of the Blue Ribbon Commission on America's Nuclear Future is not expected until January 2012, the tentative conclusions of the commission's draft report were dissected during a recent joint hearing by two subcommittees of the House of Representatives' Committee on Science, Space, and Technology. Among the more heated issues debated at the hearing was the fate of the stalled Yucca Mountain nuclear waste repository in Nevada. The Blue Ribbon Commission's (BRC) draft report includes recommendations for managing nuclear waste and for developing one or more permanent deep geological repositories and interim storage facilities, but the report does not address the future of Yucca Mountain. The BRC charter indicates that the commission is to "conduct a comprehensive review of policies for managing the back end of the nuclear fuel cycle." However, the draft report states that the commission was not asked to consider, and therefore did not address, several key issues. "We have not rendered an opinion on the suitability of the Yucca Mountain site or on the request to withdraw the license application for Yucca Mountain," the draft report states.

  4. Activity transport in nuclear generating stations

    International Nuclear Information System (INIS)

    Mitchell, A.B.

    1975-01-01

    The objective of this paper is to give a basic understanding of the operational limitations caused by radiation fields in the present design of CANDU-PHW reactors. A simple model of activity transport is described, and the significance of various radioisotopes identified. The impact which radiation fields have at the Divisional, Station Manager and Operation levels, is outlined in the context of typical work situations. (author)

  5. Heat Exchanger Design Options and Tritium Transport Study for the VHTR System

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh; Eung S. Kim

    2008-09-01

    This report presents the results of a study conducted to consider heat exchanger options and tritium transport in a very high temperature reactor (VHTR) system for the Next Generation Nuclear Plant Project. The heat exchanger options include types, arrangements, channel patterns in printed circuit heat exchangers (PCHE), coolant flow direction, and pipe configuration in shell-and-tube designs. Study considerations include: three types of heat exchanger designs (PCHE, shell-and-tube, and helical coil); single- and two-stage unit arrangements; counter-current and cross flow configurations; and straight pipes and U-tube designs in shell-and-tube type heat exchangers. Thermal designs and simple stress analyses were performed to estimate the heat exchanger options, and the Finite Element Method was applied for more detailed calculations, especially for PCHE designs. Results of the options study show that the PCHE design has the smallest volume and heat transfer area, resulting in the least tritium permeation and greatest cost savings. It is theoretically the most reliable mechanically, leading to a longer lifetime. The two-stage heat exchanger arrangement appears to be safer and more cost effective. The recommended separation temperature between first and second stages in a serial configuration is 800oC, at which the high temperature unit is about one-half the size of the total heat exchanger core volume. Based on simplified stress analyses, the high temperature unit will need to be replaced two or three times during the plant’s lifetime. Stress analysis results recommend the off-set channel pattern configuration for the PCHE because stress reduction was estimated at up to 50% in this configuration, resulting in a longer lifetime. The tritium transport study resulted in the development of a tritium behavior analysis code using the MATLAB Simulink code. In parallel, the THYTAN code, previously performed by Ohashi and Sherman (2007) on the Peach Bottom data, was revived

  6. Multipurpose containers for the transport of nuclear material: The example of transport flask CF6

    International Nuclear Information System (INIS)

    Gualdrini, G.F.; Borgia, M.G.

    1989-03-01

    The present paper summarizes the design and licensing activity carried out in the frame work of an ENEA working group which was set up with the aim of developing transport flasks for radioactive and non radioactive dangerous materials. In particular the nuclear design of the multipurpose transport flask CF6 is described. The paper was presented at the seminar on 'Nuclear wastes and transport of radioactive materials' held in Bologna on June 4th and 5th 1987 under the aegis of the Department of Physics of the University of Bologna. (author)

  7. Policies and initiatives for carbon neutrality in nordic heating and transport systems

    DEFF Research Database (Denmark)

    Muller, Jakob Glarbo; Wu, Qiuwei; Ostergaard, Jacob

    2012-01-01

    to heat pumps in the Nordic region rely on both private economic and national economic incentives. Initiatives toward carbon neutrality in the transport system are mostly concentrated on research, development and demonstration for deployment of a large number of EVs. All Nordic countries have plans......Policies and initiatives promoting carbon neutrality in the Nordic heating and transport systems are presented. The focus within heating systems is the propagation of heat pumps while the focus within transport systems is initiatives regarding electric vehicles (EVs). It is found that conversion...... for the future heating and transport systems with the ambition of realizing carbon neutrality....

  8. Currents and fluctuations of quantum heat transport in harmonic chains

    International Nuclear Information System (INIS)

    Motz, T; Ankerhold, J; Stockburger, J T

    2017-01-01

    Heat transport in open quantum systems is particularly susceptible to the modeling of system–reservoir interactions. It thus requires us to consistently treat the coupling between a quantum system and its environment. While perturbative approaches are successfully used in fields like quantum optics and quantum information, they reveal deficiencies—typically in the context of thermodynamics, when it is essential to respect additional criteria such as fluctuation-dissipation theorems. We use a non-perturbative approach for quantum dissipative dynamics based on a stochastic Liouville–von Neumann equation to provide a very general and extremely efficient formalism for heat currents and their correlations in open harmonic chains. Specific results are derived not only for first- but also for second-order moments, which requires us to account for both real and imaginary parts of bath–bath correlation functions. Spatiotemporal patterns are compared with weak coupling calculations. The regime of stronger system–reservoir couplings gives rise to an intimate interplay between reservoir fluctuations and heat transfer far from equilibrium. (paper)

  9. Transition to ballistic regime for heat transport in helium II

    Energy Technology Data Exchange (ETDEWEB)

    Sciacca, Michele, E-mail: michele.sciacca@unipa.it [Dipartimento Scienze Agrarie e Forestali, Università degli studi di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Departament de Física, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Sellitto, Antonio, E-mail: ant.sellitto@gmail.com [Dipartimento di Matematica, Informatica ed Economia, Università della Basilicata, Campus Macchia Romana, 85100 Potenza (Italy); Jou, David, E-mail: david.jou@uab.cat [Departament de Física, Universitat Autònoma de Barcelona, 08193 Bellaterra, Catalonia (Spain); Institut d' Estudis Catalans, Carme 47, 08001 Barcelona, Catalonia (Spain)

    2014-07-04

    The size-dependent and flux-dependent effective thermal conductivity of narrow capillaries filled with superfluid helium is analyzed from a thermodynamic continuum perspective. The classical Landau evaluation of the effective thermal conductivity of quiescent superfluid, or the Gorter–Mellinck regime of turbulent superfluids, is extended to describe the transition to ballistic regime in narrow channels wherein the radius R is comparable to (or smaller than) the phonon mean-free path ℓ in superfluid helium. To do so, we start from an extended equation for the heat flux incorporating non-local terms, and take into consideration a heat slip flow along the walls of the tube. This leads from an effective thermal conductivity proportional to R{sup 2} (Landau regime) to another one proportional to Rℓ (ballistic regime). We consider two kinds of flows: along cylindrical pipes and along two infinite parallel plates. - Highlights: • Heat transport in counterflow helium in the ballistic regime. • The one-fluid model based on the Extended Thermodynamics is used. • The transition from the Landau regime to the ballistic regime. • The transition from quantum turbulence to ballistic regime.

  10. Currents and fluctuations of quantum heat transport in harmonic chains

    Science.gov (United States)

    Motz, T.; Ankerhold, J.; Stockburger, J. T.

    2017-05-01

    Heat transport in open quantum systems is particularly susceptible to the modeling of system-reservoir interactions. It thus requires us to consistently treat the coupling between a quantum system and its environment. While perturbative approaches are successfully used in fields like quantum optics and quantum information, they reveal deficiencies—typically in the context of thermodynamics, when it is essential to respect additional criteria such as fluctuation-dissipation theorems. We use a non-perturbative approach for quantum dissipative dynamics based on a stochastic Liouville-von Neumann equation to provide a very general and extremely efficient formalism for heat currents and their correlations in open harmonic chains. Specific results are derived not only for first- but also for second-order moments, which requires us to account for both real and imaginary parts of bath-bath correlation functions. Spatiotemporal patterns are compared with weak coupling calculations. The regime of stronger system-reservoir couplings gives rise to an intimate interplay between reservoir fluctuations and heat transfer far from equilibrium.

  11. Feasibility study of a dedicate nuclear desalination system: Low-pressure inherent heat sink nuclear desalination plant (LIND)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ho Sik; No, Hee Cheon; Jo, Yu Gwan; Wivisono, Andhika Feri; Park, Byung Ha; Choi, Jin Young; Lee, Jeong Ik; Jeong, Yong Hoon; Cho, Nam Zin [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-04-15

    In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MW{sub th} and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

  12. Feasibility study of a dedicated nuclear desalination system: Low-pressure Inherent heat sink Nuclear Desalination plant (LIND

    Directory of Open Access Journals (Sweden)

    Ho Sik Kim

    2015-04-01

    Full Text Available In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal–hydraulic and neutronic design requirements. In a thermal–hydraulic analysis using an analytical method based on the Wooton–Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MWth and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

  13. Vulnerability Analysis Considerations for the Transportation of Special Nuclear Material

    Energy Technology Data Exchange (ETDEWEB)

    Nicholson, Lary G.; Purvis, James W.

    1999-07-21

    The vulnerability analysis methodology developed for fixed nuclear material sites has proven to be extremely effective in assessing associated transportation issues. The basic methods and techniques used are directly applicable to conducting a transportation vulnerability analysis. The purpose of this paper is to illustrate that the same physical protection elements (detection, delay, and response) are present, although the response force plays a dominant role in preventing the theft or sabotage of material. Transportation systems are continuously exposed to the general public whereas the fixed site location by its very nature restricts general public access.

  14. Vulnerability Analysis Considerations for the Transportation of Special Nuclear Material

    International Nuclear Information System (INIS)

    Nicholson, Lary G.; Purvis, James W.

    1999-01-01

    The vulnerability analysis methodology developed for fixed nuclear material sites has proven to be extremely effective in assessing associated transportation issues. The basic methods and techniques used are directly applicable to conducting a transportation vulnerability analysis. The purpose of this paper is to illustrate that the same physical protection elements (detection, delay, and response) are present, although the response force plays a dominant role in preventing the theft or sabotage of material. Transportation systems are continuously exposed to the general public whereas the fixed site location by its very nature restricts general public access

  15. Assembly for transport and storage of radioactive nuclear fuel elements

    International Nuclear Information System (INIS)

    Myers, G.

    1978-01-01

    The invention concerns the self-control of coolant deficiencies on the transport of spent fuel elements from nuclear reactors. It guarantees that drying out of the fuel elements is prevented in case of a change of volume of the fluid contained in storage tanks and accumulators and serving as coolant and shielding medium. (TK) [de

  16. Studies and research concerning BNFP. Nuclear spent fuel transportation studies

    International Nuclear Information System (INIS)

    Anderson, R.T.; Maier, J.B.

    1979-11-01

    Currently, there are a number of institutional problems associated with the shipment of spent fuel assemblies from commercial nuclear power plants: new and conflicting regulations, embargoing of certain routes, imposition of transport safeguards, physical security in-transit, and a lack of definition of when and where the fuel will be moved. This report presents a summary of these types and kinds of problems. It represents the results of evaluations performed relative to fuel receipt at the Barnwell Nuclear Fuel Plant. Case studies were made which address existing reactor sites with near-term spent fuel transportation needs. Shipment by either highway, rail, water, or intermodal water-rail was considered. The report identifies the impact of new regulations and uncertainty caused by indeterminate regulatory policy and lack of action on spent fuel acceptance and storage. This stagnant situation has made it impossible for industry to determine realistic transportation scenarios for business planning and financial risk analysis. A current lack of private investment in nuclear transportation equipment is expected to further prolong the problems associated with nuclear spent fuel and waste disposition. These problems are expected to intensify in the 1980's and in certain cases will make continuing reactor plant operation difficult or impossible

  17. Program strategy document for the Nuclear Materials Transportation Technology Center

    International Nuclear Information System (INIS)

    Jefferson, R.M.

    1979-07-01

    A multiyear program plan is presented which describes the program of the Nuclear Materials Transportation Technology Center (TIC) at Sandia Laboratories. The work element plans, along with their corresponding work breakdown structures, are presented for TTC activities in the areas of Technology and Information Center, Systems Development, Technology, and Institutional Issues for the years from 1979 to 1985

  18. Dissecting the signaling events that impact classical nuclear import and target nuclear transport factors.

    Directory of Open Access Journals (Sweden)

    Mohamed Kodiha

    2009-12-01

    Full Text Available Signaling through MEK-->ERK1/2 and PI3 kinases is implicated in many aspects of cell physiology, including the survival of oxidant exposure. Oxidants play a role in numerous physiological and pathophysiological processes, many of which rely on transport in and out of the nucleus. However, how oxidative stress impacts nuclear trafficking is not well defined.To better understand the effect of stress on nucleocytoplasmic trafficking, we exposed cells to the oxidant diethyl maleate. This treatment activated MEK-->ERK1/2 as well as PI3 kinase-->Akt cascades and triggered the inhibition of classical nuclear import. To define the molecular mechanisms that regulate nuclear transport, we examined whether MEK and PI3 kinase signaling affected the localization of key transport factors. Using recently developed tools for image acquisition and analysis, the subcellular distributions of importin-alpha, CAS, and nucleoporins Nup153 and Nup88 were quantified in different cellular compartments. These studies identified specific profiles for the localization of transport factors in the nucleus and cytoplasm, and at the nuclear envelope. Our results demonstrate that MEK and PI3 kinase signaling as well as oxidative stress control nuclear trafficking and the localization of transport components. Furthermore, stress not only induced changes in transport factor distribution, but also upregulated post-translational modification of transport factors. Our results are consistent with the idea that the phosphorylation of importin-alpha, CAS, Nup153, and Nup88, and the O-GlcNAc modification of Nup153 increase when cells are exposed to oxidant.Our studies defined the complex regulation of classical nuclear import and identified key transport factors that are targeted by stress, MEK, and PI3 kinase signaling.

  19. Reactor type choice and characteristics for a small nuclear heat and electricity co-generation plant

    International Nuclear Information System (INIS)

    Liu Kukui; Li Manchang; Tang Chuanbao

    1997-01-01

    In China heat supply consumes more than 70 percent of the primary energy resource, which makes for heavy traffic and transportation and produces a lot of polluting materials such as NO x , SO x and CO 2 because of use of the fossil fuel. The utilization of nuclear power into the heat and electricity co-generation plant contributes to the global environmental protection. The basic concept of the nuclear system is an integral type reactor with three circuits. The primary circuit equipment is enclosed in and linked up directly with reactor vessel. The third circuit produces steam for heat and electricity supply. This paper presents basic requirements, reactor type choice, design characteristics, economy for a nuclear co-generation plant and its future application. The choice of the main parameters and the main technological process is the key problem of the nuclear plant design. To make this paper clearer, take for example a double-reactor plant of 450 x 2MW thermal power. There are two sorts of main technological processes. One is a water-water-steam process. Another is water-steam-steam process. Compared the two sorts, the design which adopted the water-water-steam technological process has much more advantage. The system is simplified, the operation reliability is increased, the primary pressure reduces a lot, the temperature difference between the secondary and the third circuits becomes larger, so the size and capacity of the main components will be smaller, the scale and the cost of the building will be cut down. In this design, the secondary circuit pressure is the highest among that of the three circuits. So the primary circuit radioactivity can not leak into the third circuit in case of accidents. (author)

  20. High Efficiency RF Heating for Small Nuclear Fusion Rocket Engines, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — High power nuclear fusion propulsion systems will require high efficiency radio-frequency heating systems in the MHz range for plasma heating. This proposal is for a...

  1. NUCLEAR HEATING IN LIF DOSEMETERS IN A FUSION NEUTRON FIELD, TRIAL OF DIRECT COMPARISON OF EXPERIMENTAL AND SIMULATED RESULTS.

    Science.gov (United States)

    Pohorecki, Wladyslaw; Obryk, Barbara

    2017-09-29

    The results of nuclear heating measured by means of thermoluminescent dosemeters (TLD-LiF) in a Cu block irradiated by 14 MeV neutrons are presented. The integral Cu experiment relevant for verification of copper nuclear data at neutron energies characteristic for fusion facilities was performed in the ENEA FNG Laboratory at Frascati. Five types of TLDs were used: highly photon sensitive LiF:Mg,Cu,P (MCP-N), 7LiF:Mg,Cu,P (MCP-7) and standard, lower sensitivity LiF:Mg,Ti (MTS-N), 7LiF:Mg,Ti (MTS-7) and 6LiF:Mg,Ti (MTS-6). Calibration of the detectors was performed with gamma rays in terms of air-kerma (10 mGy of 137Cs air-kerma). Nuclear heating in the Cu block was also calculated with the use of MCNP transport code Nuclear heating in Cu and air in TLD's positions was calculated as well. The nuclear heating contribution from all simulated by MCNP6 code particles including protons, deuterons, alphas tritons and heavier ions produced by the neutron interactions were calculated. A trial of the direct comparison between experimental results and results of simulation was performed. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  2. Simulation of thermal behavior of nuclear fuel rod by electrically heated pin

    International Nuclear Information System (INIS)

    Carajilescov, P.

    1985-01-01

    The utilization of electrically heated rods for the simulation of nuclear fuel rods represents an universally adopted method by the nuclear industry to study thermalhydraulic problems. The present work represents the development of a method to obtain the time variation of the electric linear power necessary to simulate a given nuclear power transient in order to yield the same temperature and heat flux conditions in the surface of the electrical heater that would be obtained by the nuclear fuel rod. (Author) [pt

  3. Experimental study on the supercritical startup and heat transport capability of a neon-charged cryogenic loop heat pipe

    International Nuclear Information System (INIS)

    Guo, Yuandong; Lin, Guiping; He, Jiang; Bai, Lizhan; Zhang, Hongxing; Miao, Jianyin

    2017-01-01

    Highlights: • A neon-charged CLHP integrated with a G-M cryocooler was designed and investigated. • The CLHP can realize the supercritical startup with an auxiliary heat load of 1.5 W. • Maximum heat transport capability of the CLHP was 4.5 W over a distance of 0.6 m. • There existed an optimum auxiliary heat load to expedite the supercritical startup. • There existed an optimum charged pressure to reach the largest heat transfer limit. - Abstract: Neon-charged cryogenic loop heat pipe (CLHP) can realize efficient cryogenic heat transport in the temperature range of 30–40 K, and promises great application potential in the thermal control of future space infrared exploration system. In this work, extensive experimental studies on the supercritical startup and heat transport capability of a neon-charged CLHP integrated with a G-M cryocooler were carried out, where the effects of the auxiliary heat load applied to the secondary evaporator and charged pressure of the working fluid were investigated. Experimental results showed that the CLHP could successfully realize the supercritical startup with an auxiliary heat load of 1.5 W, and there existed an optimum auxiliary heat load and charged pressure of the working fluid respectively, to achieve the maximum temperature drop rate of the primary evaporator during the supercritical startup. The CLHP could reach a maximum heat transport capability of 4.5 W over a distance of 0.6 m corresponding to the optimum charged pressure of the working fluid; however, the heat transport capability decreased with the increase of the auxiliary heat load. Furthermore, the inherent mechanisms responsible for the phenomena observed in the experiments were analyzed and discussed, to provide a better understanding from the theoretical view.

  4. Development of CANDU 6 Primary Heat Transport System Modeling Program

    International Nuclear Information System (INIS)

    Seo, Hyung-beom; Kim, Sung-min; Park, Joong-woo; Kim, Kwang-su; Ko, Dae-hack; Han, Bong-seob

    2007-01-01

    NUCIRC is a steady-state thermal-hydraulic code used for design and performance analyses of CANDU Heat Transport System. The code is used to build PHT model in Wolsong NPP and to calculate channel flow distribution. Wolsong NPP has to calculate channel flow distribution and quality of coolant at the ROH header after every outage by OPP (Operating Policy and Principal). PHT modeling work is time consuming which need a lot of operation experience and specialty. It is very difficult to build PHT model as plant operator in two weeks which is obligate for plant operation after every outage. That is why Wolsong NPP develop NUMODEL (NUcirc MODELing) with many-years experience and a know-how of using NUCIRC code. NUMODEL is computer program which is used to create PHT model based on utilizing NUCIRC code

  5. Periodic inspection for safety of CANDU heat transport piping systems

    International Nuclear Information System (INIS)

    Ellyin, F.

    1979-10-01

    Periodic inspection of heat transport and emergency core cooling piping systems is intended to maintain an adequate level of safety throughout the life of the plant, and to protect plant personnel and the public from the consequences of a failure and release of fission products. This report outlines a rational approach to the periodic inspection based on a fully probabilistic model. It demonstrates the methodology based on theoretical treatment and experimental data whereby the strength of a pressurized pipe or vessel containing a defect could be evaluated. It also shows how the extension of the defect at various lifetimes could be predicted. These relationships are prerequisite for the probabilistic formulation and analysis for the periodic inspection of piping systems

  6. Optimal wall spacing for heat transport in thermal convection

    Energy Technology Data Exchange (ETDEWEB)

    Shishkina, Olga [Max Planck Institute for Dynamics and Self-Organization, Goettingen (Germany)

    2016-11-01

    The simulation of RB flow for Ra up to 1 x 10{sup 10} is computationally expensive in terms of computing power and hard disk storage. Thus, we gratefully acknowledge the computational resources supported by Leibniz-Rechenzentrum Munich. Compared to Γ=1 situation, a new physical picture of heat transport is identified here at Γ{sub opt} for any explored Ra. Therefore, a detailed comparison between Γ=1 and Γ=Γ{sub opt} is valuable for our further research, for example, their vertical temperature and velocity profiles. Additionally, we plan to compare the fluid with different Pr under geometrical confinement, which are computationally expensive for the situations of Pr<<1 and Pr>>1.

  7. Next generation CANDU heat transport system parameter assessment

    International Nuclear Information System (INIS)

    Hau, K.F.; Love, J.W.; Vadera, M.; Vecchiarelli, J.

    2001-01-01

    AECL has initiated an innovative program to develop the next generation of technologies for CANDU reactors, and to apply them to a highly cost-effective new family of next generation power plants. Four major design changes were considered in the present conceptual design of the Heat Transport System (HTS) for the Next Generation (NG) CANDU. These include: light water replacement of heavy water as coolant, a compact core design resulting from a fuel channel lattice pitch reduction, use of Slightly Enriched Uranium (SEU) CANFLEX fuel bundles, and higher HTS and Turbine Generator (TG) operating pressures and temperatures. In designing the HTS, the goal is to reduce the capital cost while meeting the design and safety requirements with robust safety margins. This paper describes the studies to optimize key HTS parameters, including the assessment methodology and the basis of proposed design conditions for the NG CANDU HTS. (author)

  8. Matching design and layout of the nuclear heating station AST-500 to the district heating network conditions in the GDR

    International Nuclear Information System (INIS)

    Luetzow, K.; Bordihn, S.

    1988-01-01

    Nuclear heating stations provide one possibility of supplying nuclear heat. They are uniquely intended for heat supply. The well-known Soviet nuclear heating station AST-500 with an output temperature of 150 0 C in the network pipes is below the value designed for some heat supply systems in the GDR. On the basis of a general method applicable to the choice of optimum parameters, proposals are placed for optimally matching design and layout of the AST-500 station to higher output temperatures. Minimization of the difference in expenses between the modified version and the well-known basic version of the AST-500 station has been used as the objective function of optimization. (author)

  9. Heat transport modeling of the dot spectroscopy platform on NIF

    Science.gov (United States)

    Farmer, W. A.; Jones, O. S.; Barrios, M. A.; Strozzi, D. J.; Koning, J. M.; Kerbel, G. D.; Hinkel, D. E.; Moody, J. D.; Suter, L. J.; Liedahl, D. A.; Lemos, N.; Eder, D. C.; Kauffman, R. L.; Landen, O. L.; Moore, A. S.; Schneider, M. B.

    2018-04-01

    Electron heat transport within an inertial-fusion hohlraum plasma is difficult to model due to the complex interaction of kinetic plasma effects, magnetic fields, laser-plasma interactions, and microturbulence. Here, simulations using the radiation-hydrodynamic code, HYDRA, are compared to hohlraum plasma experiments which contain a Manganese-Cobalt tracer dot (Barrios et al 2016 Phys. Plasmas 23 056307). The dot is placed either on the capsule or on a film midway between the capsule and the laser-entrance hole. From spectroscopic measurements, electron temperature and position of the dot are inferred. Simulations are performed with ad hoc flux limiters of f = 0.15 and f = 0.03 (with electron heat flux, q, limited to fnT 3/2/m 1/2), and two more physical means of flux limitation: the magnetohydrodynamics and nonlocal packages. The nonlocal model agrees best with the temperature of the dot-on-film and dot-on-capsule. The hohlraum produced x-ray flux is over-predicted by roughly ˜11% for the f = 0.03 model and the remaining models by ˜16%. The simulated trajectories of the dot-on-capsule are slightly ahead of the experimental trajectory for all but the f = 0.03 model. The simulated dot-on-film position disagrees with the experimental measurement for all transport models. In the MHD simulation of the dot-on-film, the dot is strongly perturbative, though the simulation predicts a peak dot-on-film temperature 2-3 keV higher than the measurement. This suggests a deficiency in the MHD modeling possibly due to the neglect of the Righi-Leduc term or interpenetrating flows of multiple ion species which would reduce the strength of the self-generated fields.

  10. Modeling studies of multiphase fluid and heat flow processes in nuclear waste isolation

    International Nuclear Information System (INIS)

    Pruess, K.

    1989-01-01

    Multiphase fluid and heat flow plays an important role in many problems relating to the disposal of nuclear wastes in geologic media. Examples include boiling and condensation processes near heat-generating wastes, flow of water and formation gas in partially saturated formations, evolution of a free gas phase from waste package corrosion in initially water-saturated environments, and redistribution (dissolution, transport and precipitation) of rock minerals in non-isothermal flow fields. Such processes may strongly impact upon waste package and repository design considerations and performance. This paper summarizes important physical phenomena occurring in multiphase and nonisothermal flows, as well as techniques for their mathematical modeling and numerical simulation. Illustrative applications are given for a number of specific fluid and heat flow problems, including: thermohydrologic conditions near heat-generating waste packages in the unsaturated zone; repositorywide convection effects in the unsaturated zone; effects of quartz dissolution and precipitation for disposal in the saturated zone; and gas pressurization and flow effects from corrosion of low-level waste packages

  11. Modeling studies for multiphase fluid and heat flow processes in nuclear waste isolation

    International Nuclear Information System (INIS)

    Pruess, K.

    1988-07-01

    Multiphase fluid and heat flow plays an important role in many problems relating to the disposal of nuclear wastes in geologic media. Examples include boiling and condensation processes near heat-generating wastes, flow of water and formation gas in partially saturated formations, evolution of a free gas phase from waste package corrosion in initially water-saturated environments, and redistribution (dissolution, transport, and precipitation) of rock minerals in non-isothermal flow fields. Such processes may strongly impact upon waste package and repository design considerations and performance. This paper summarizes important physical phenomena occurring in multiphase and nonisothermal flows, as well as techniques for their mathematical modeling and numerical simulation. Illustrative applications are given for a number of specific fluid and heat flow problems, including: thermohydrologic conditions near heat-generating waste packages in the unsaturated zone; repository-wide convection effects in the unsaturated zone; effects of quartz dissolution and precipitation for disposal in the saturated zone; and gas pressurization and flow corrosion of low-level waste packages. 34 refs; 7 figs; 2 tabs

  12. Transportation of nuclear materials: the nuclear focus of the 80's

    International Nuclear Information System (INIS)

    Meyers, S.; Hardin, E.C. Jr.; Jefferson, R.M.

    1980-01-01

    The transport of radioactive material has been carried out since the inception of the nuclear age (over 30 years) with an unparralled safety record. Despite these achievements, there is a need to strive for improvements, to develop safer and more efficient transportation systems, moreover to perform these tasks in a highly visible manner so that public concern can be allayed. But, in the same vein that the past record is not of itself sufficient, neither is public participation the solution to all the issues surrounding the transportation of radioactive materials. The solutions to the problems facing the nuclear transport industry involve many disciplines, much of which rest on a foundation of sound technology. This conference is built around a core of papers on the developing technology of nuclear transportation: on systems, their design and development, their manufacturing processes, their operation and the methodologies of quality assurance in each of these activities. The role of IAEA in the collecting of data to compile information on the flow of radioactive materials, the mode of transport and the corresponding accident/incident experience, as well as its role in initiating a program to develop a worldwide uniform methodology to address the risks of transporting radioactive materials are covered in this symposium

  13. Intercontinental nuclear transport from the private international law perspective

    International Nuclear Information System (INIS)

    Magnus, U.

    2000-01-01

    The aim of this paper is to give a survey on choice of law rules which apply outside the nuclear liability conventions in case of damage caused by international nuclear transports. We found a remarkable variety of solutions. Some of the solutions make it difficult or even impossible to predict in advance which substantive law in a hypothetical case would apply. These difficulties are increased by the fact that more often than not, a victim can choose where to sue and thereby also influence the final outcome of a case. As far as private international law rules apply - and as mentioned the non-ratification of the nuclear liability conventions by many nuclear states forces us to fall back on the choice of law rules in many cases - the applicable law and the hypothetical level of compensation therefore often remain uncertain when judged at the time of organisation of the nuclear transport. However, at this time the question of undertaking risks and of insurability must be decided. (author)

  14. Climate in the absence of ocean heat transport

    Science.gov (United States)

    Rose, B. E. J.

    2017-12-01

    The energy transported by the oceans to mid- and high latitudes is small compared to the atmosphere, yet exerts an outsized influence on climate. A key reason is the strong interaction between ocean heat transport (OHT) and sea ice extent. I quantify the absolute climatic impact of OHT using the state-of-the-art CESM simulations by comparing a realistic control climate against a slab ocean simulation in which OHT is disabled. The absence of OHT leads to a massive expansion of sea ice into the subtropics in both hemispheres, and a 24 K global cooling. Analysis of the transient simulation after setting the OHT to zero reveals a global cooling process fueled by a runaway sea ice albedo feedback. This process is eventually self-limiting in the cold climate due to a combination of subtropical cloud feedbacks and surface wind effects that are both connected to a massive spin-up of the atmospheric Hadley circulation. A parameter sensitivity study shows that the simulated climate is far more sensitive to small changes in ice surface albedo in the absence of OHT. I conclude that the oceans are responsible for an enormous global warming by mitigating an otherwise very potent sea ice albedo feedback, but that the magnitude of this effect is rather uncertain. These simulations provide a graphic illustration of how the intimate coupling between sea ice and ocean circulation governs the present-day climate, and by extension, highlight the importance of modeling ocean - sea ice interaction with high fidelity.

  15. Experimental simulation study on hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Zhang Youjie; Jia Haijun; Bo Jinhai; Hong Liuming; Bo Hanliang; Liu Zhiyong

    1997-07-01

    The hydraulic behavior of the main heat exchanger of Daqing 200 MW nuclear heating reactor is studied through a 1:2.33 test model. The design and other feature of the test model is described. The experimental results show that the flow resistance coefficient of the heat exchanger becomes self-simulation when Reynolds number is greater than 5000. The value of flow resistance coefficient at self-simulation condition and the distribution of pressure drop in the heat exchanger are given through experiment. The option design to reduce flow resistance is proposed. The designed and experimental value for the flow resistance coefficient are in good agreement. The variation of system parameters during flow excursion was described. The experimental results are of great significant for the final design of the main heat exchanger of Daqing 200 MW nuclear heating reactor. (2 refs., 5 figs., 1 tab.)

  16. Multiphase, multicomponent flow and transport models for Nuclear Test-Ban Treaty monitoring and nuclear waste disposal applications

    Science.gov (United States)

    Jordan, Amy

    Open challenges remain in using numerical models of subsurface flow and transport systems to make useful predictions related to nuclear waste storage and nonproliferation. The work presented here addresses the sensitivity of model results to unknown parameters, states, and processes, particularly uncertainties related to incorporating previously unrepresented processes (e.g., explosion-induced fracturing, hydrous mineral dehydration) into a subsurface flow and transport numerical simulator. The Finite Element Heat and Mass (FEHM) transfer code is used for all numerical models in this research. An experimental campaign intended to validate the predictive capability of numerical models that include the strongly coupled thermal, hydrological, and chemical processes in bedded salt is also presented. Underground nuclear explosions (UNEs) produce radionuclide gases that may seep to the surface over weeks to months. The estimated timing of gas arrival at the surface may be used to deploy personnel and equipment to the site of a suspected UNE, if allowed under the terms of the Comprehensive Nuclear Test-Ban Treaty. A model was developed using FEHM that considers barometrically pumped gas transport through a simplified fractured medium and was used to quantify the impact of uncertainties in hydrologic parameters (fracture aperture, matrix permeability, porosity, and saturation) and season of detonation on the timing of gas breakthrough. Numerical sensitivity analyses were performed for the case of a 1 kt UNE at a 400 m burial depth. Gas arrival time was found to be most affected by matrix permeability and fracture aperture. Gases having higher diffusivity were more sensitive to uncertainty in the rock properties. The effect of seasonality in the barometric pressure forcing was found to be important, with detonations in March the least likely to be detectable based on barometric data for Rainier Mesa, Nevada. Monte Carlo modeling was also used to predict the window of

  17. After heat distribution of a mobile nuclear power plant

    Science.gov (United States)

    Parker, W. G.; Vanbibber, L. E.; Tang, Y. S.

    1971-01-01

    A computer program was developed to analyze the transient afterheat temperature and pressure response of a mobile gas-cooled reactor power plant following impact. The program considers (in addition to the standard modes of heat transfer) fission product decay and transport, metal-water reactions, core and shield melting and displacement, and pressure and containment vessel stress response. Analyses were performed for eight cases (both deformed and undeformed models) to verify operability of the program options. The results indicated that for a 350 psi (241 n/sq cm) initial internal pressure, the containment vessel can survive over 100,000 seconds following impact before creep rupture occurs. Recommendations were developed as to directions for redesign to extend containment vessel life.

  18. Experiences in certification of packages for transportation of fresh nuclear fuel in the context of new safety requirements established by IAEA regulations (IAEA-96 regulations, ST-1) for air transportation of nuclear materials (requirements to C-type packages)

    Energy Technology Data Exchange (ETDEWEB)

    Dudai, V.I.; Kovtun, A.D.; Matveev, V.Z.; Morenko, A.I.; Nilulin, V.M.; Shapovalov, V.I.; Yakushev, V.A.; Bobrovsky, V.S.; Rozhkov, V.V.; Agapov, A.M.; Kolesnikov, A.S. [Russian Federal Nuclear Centre - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)]|[JSC ' ' MSZ' ' , Electrostal (Russian Federation)]|[JSC ' ' NPCC' ' , Novosibirsk (Russian Federation)]|[Minatom of Russia, Moscow (Russian Federation)]|[Gosatomnadzor of Russia, Moscow (Russian Federation)

    2004-07-01

    Every year in Russia, a large amount of domestic and international transportation of fresh nuclear fuel (FNF) used in Russian and foreign energy and research atomic reactors and referred to fissile materials based on IAEA Regulations is performed. Here, bulk transportation is performed by air, and it concerns international transportation in particular. According to national ''Main Regulations for Safe Transport and physical Protection of Nuclear Materials (OPBZ- 83)'' and ''Regulations for the Safe Transport of Radioactive Materials'' of the International Atomic Energy Agency (IAEA Regulations), nuclear and radiation security under normal (accident free) and accident conditions of transport must be completely provided by the package design. In this context, high requirements to fissile packages exposed to heat and mechanical loads in transport accidents are imposed. A long-standing experience in accident free transportation of FM has shown that such approach to provide nuclear and radiation security pays for itself completely. Nevertheless, once in 10 years the International Atomic Energy Agency on every revision of the ''Regulations for the Safe Transport of Radioactive Materials'' places more stringent requirements upon the FM and transportation thereof, resulting from the objectively increasing risk associated with constant rise in volume and density of transportation, and also strained social and economical situation in a number of regions in the world. In the new edition of the IAEA Regulations (ST-1), published in 1996 and brought into force in 2001 (IAEA-96 Regulations), the requirements to FM packages conveyed by aircraft were radically changed. These requirements are completely presented in new Russian ''Regulations for the Safe Transport of Radioactive Materials'' (PBTRM- 2004) which will be brought into force in the time ahead.

  19. Transport code and nuclear data in intermediate energy region

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

    1998-11-01

    We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

  20. 75 FR 38151 - Governors' Designees Receiving Advance Notification of Transportation of Nuclear Waste

    Science.gov (United States)

    2010-07-01

    ... Transportation of Nuclear Waste On January 6, 1982 (47 FR 596 and 47 FR 600), the U.S. Nuclear Regulatory... their designees by NRC licensees prior to transportation of certain shipments of nuclear waste and spent... notification for Part 71 is for large quantity shipments of radioactive waste (and of spent nuclear reactor...

  1. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  2. Dynamics of miRNA biogenesis and nuclear transport

    Directory of Open Access Journals (Sweden)

    Kotipalli Aneesh

    2016-12-01

    Full Text Available MicroRNAs (miRNAs are short noncoding RNA sequences ~22 nucleotides in length that play an important role in gene regulation-transcription and translation. The processing of these miRNAs takes place in both the nucleus and the cytoplasm while the final maturation occurs in the cytoplasm. Some mature miRNAs with nuclear localisation signals (NLS are transported back to the nucleus and some remain in the cytoplasm. The functional roles of these miRNAs are seen in both the nucleus and the cytoplasm. In the nucleus, miRNAs regulate gene expression by binding to the targeted promoter sequences and affect either the transcriptional gene silencing (TGS or transcriptional gene activation (TGA. In the cytoplasm, targeted mRNAs are translationally repressed or cleaved based on the complementarity between the two sequences at the seed region of miRNA and mRNA. The selective transport of mature miRNAs to the nucleus follows the classical nuclear import mechanism. The classical nuclear import mechanism is a highly regulated process, involving exportins and importins. The nuclear pore complex (NPC regulates all these transport events like a gate keeper. The half-life of miRNAs is rather low, so within a short time miRNAs perform their function. Temporal studies of miRNA biogenesis are, therefore, useful. We have carried out simulation studies for important miRNA biogenesis steps and also classical nuclear import mechanism using ordinary differential equation (ODE solver in the Octave software.

  3. Dynamics of miRNA biogenesis and nuclear transport.

    Science.gov (United States)

    Kotipalli, Aneesh; Gutti, Ravikumar; Mitra, Chanchal K

    2016-12-22

    MicroRNAs (miRNAs) are short noncoding RNA sequences ~22 nucleotides in length that play an important role in gene regulation-transcription and translation. The processing of these miRNAs takes place in both the nucleus and the cytoplasm while the final maturation occurs in the cytoplasm. Some mature miRNAs with nuclear localisation signals (NLS) are transported back to the nucleus and some remain in the cytoplasm. The functional roles of these miRNAs are seen in both the nucleus and the cytoplasm. In the nucleus, miRNAs regulate gene expression by binding to the targeted promoter sequences and affect either the transcriptional gene silencing (TGS) or transcriptional gene activation (TGA). In the cytoplasm, targeted mRNAs are translationally repressed or cleaved based on the complementarity between the two sequences at the seed region of miRNA and mRNA. The selective transport of mature miRNAs to the nucleus follows the classical nuclear import mechanism. The classical nuclear import mechanism is a highly regulated process, involving exportins and importins. The nuclear pore complex (NPC) regulates all these transport events like a gate keeper. The half-life of miRNAs is rather low, so within a short time miRNAs perform their function. Temporal studies of miRNA biogenesis are, therefore, useful. We have carried out simulation studies for important miRNA biogenesis steps and also classical nuclear import mechanism using ordinary differential equation (ODE) solver in the Octave software.

  4. Experience of air transport of nuclear fuel material as type A package

    International Nuclear Information System (INIS)

    Kawasaki, Masashi; Kageyama, Tomio; Suzuki, Toru

    2004-01-01

    Special law on nuclear disaster countermeasures (hereafter called as to nuclear disaster countermeasures low) that is domestic law for dealing with measures for nuclear disaster, was enforced in June, 2000. Therefore, nuclear enterprise was obliged to report accidents as required by nuclear disaster countermeasures law, besides meeting the technical requirement of existent transport regulation. For overseas procurement of plutonium reference materials that are needed for material accountability, A Type package must be transported by air. Therefore, concept of air transport of nuclear fuel materials according to the nuclear disaster countermeasures law was discussed, and the manual including measures against accident in air transport was prepared for the oversea procurement. In this presentation, the concept of air transport of A Type package containing nuclear fuel materials according to the nuclear disaster countermeasures law, and the experience of a transportation of plutonium solution from France are shown. (author)

  5. Heat transport analysis in a district heating and snow melting system in Sapporo and Ishikari, Hokkaido applying waste heat from GTHTR300

    International Nuclear Information System (INIS)

    Kasahara, Seiji; Kamiji, Yu; Terada, Atsuhiko; Yan Xing; Inagaki, Yoshiyuki; Murata, Tetsuya; Mori, Michitsugu

    2015-01-01

    A district heating and snow melting system utilizing waste heat from Gas Turbine High temperature Gas Reactor of 300 MW e (GTHTR300), a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. Application areas are set in Sapporo and Ishikari, the heavy snowfall cities in Northern Japan. The heat transport analyses are carried out by modeling the components in the system; pipelines of the secondary water loops between GTHTR300s and heat demand district and heat exchangers to transport the heat from the secondary water loops to the tertiary loops in the district. Double pipe for the secondary loops are advantageous for less heat loss and smaller excavation area. On the other hand, these pipes has disadvantage of more electricity consumption for pumping. Most of the heat demand in the month of maximum requirement can be supplied by 2 GTHTR300s and delivered by 9 secondary loops and around 5000 heat exchangers. Closer location of GTHTR300 site to the heat demand district is largely advantageous economically. Less decrease of the distance from 40 km to 20 km made the heat loss half and cost of the heat transfer system 22% smaller. (author)

  6. A Preliminary Evaluation of Using Fill Materials to Stabilize Used Nuclear Fuel During Storage and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Lahti, Erik A.; Richmond, David J.

    2012-08-01

    This report contains a preliminary evaluation of potential fill materials that could be used to fill void spaces in and around used nuclear fuel contained in dry storage canisters in order to stabilize the geometry and mechanical structure of the used nuclear fuel during extended storage and transportation after extended storage. Previous work is summarized, conceptual descriptions of how canisters might be filled were developed, and requirements for potential fill materials were developed. Elements of the requirements included criticality avoidance, heat transfer or thermodynamic properties, homogeneity and rheological properties, retrievability, material availability and cost, weight and radiation shielding, and operational considerations. Potential fill materials were grouped into 5 categories and their properties, advantages, disadvantages, and requirements for future testing were discussed. The categories were molten materials, which included molten metals and paraffin; particulates and beads; resins; foams; and grout. Based on this analysis, further development of fill materials to stabilize used nuclear fuel during storage and transportation is not recommended unless options such as showing that the fuel remains intact or canning of used nuclear fuel do not prove to be feasible.

  7. Nuclear Energy R and D Imperative 3: Enable a Transition Away from Fossil Fuel in the Transportation and Industrial Sectors

    International Nuclear Information System (INIS)

    Petti, David; Herring, J. Stephen

    2010-01-01

    As described in the Department of Energy Office of Nuclear Energy's Nuclear Energy R and D Roadmap, nuclear energy can play a significant role in supplying energy for a growing economy while reducing both our dependence on foreign energy supplies and emissions from the burning of fossil fuels. The industrial and transportation sectors are responsible for more than half of the greenhouse gas emissions in the U.S., and imported oil supplies 70% of the energy used in the transportation sector. It is therefore important to examine the various ways nuclear energy can facilitate a transition away from fossil fuels to secure environmentally sustainable production and use of energy in the transportation and manufacturing industry sectors. Imperative 3 of the Nuclear Energy R and D Roadmap, entitled 'Enable a Transition Away from Fossil Fuels by Producing Process Heat for use in the Transportation and Industrial Sectors', addresses this need. This document presents an Implementation Plan for R and D efforts related to this imperative. The expanded use of nuclear energy beyond the electrical grid will contribute significantly to overcoming the three inter-linked energy challenges facing U.S. industry: the rising and volatile prices for premium fossil fuels such as oil and natural gas, dependence on foreign sources for these fuels, and the risks of climate change resulting from carbon emissions. Nuclear energy could be used in the industrial and transportation sectors to: (1) Generate high temperature process heat and electricity to serve industrial needs including the production of chemical feedstocks for use in manufacturing premium fuels and fertilizer products, (2) Produce hydrogen for industrial processes and transportation fuels, and (3) Provide clean water for human consumption by desalination and promote wastewater treatment using low-grade nuclear heat as a useful additional benefit. Opening new avenues for nuclear energy will significantly enhance our nation

  8. Nuclear Energy R&D Imperative 3: Enable a Transition Away from Fossil Fuel in the Transportation and Industrial Sectors

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; J. Stephen Herring

    2010-03-01

    As described in the Department of Energy Office of Nuclear Energy’s Nuclear Energy R&D Roadmap, nuclear energy can play a significant role in supplying energy for a growing economy while reducing both our dependence on foreign energy supplies and emissions from the burning of fossil fuels. The industrial and transportation sectors are responsible for more than half of the greenhouse gas emissions in the U.S., and imported oil supplies 70% of the energy used in the transportation sector. It is therefore important to examine the various ways nuclear energy can facilitate a transition away from fossil fuels to secure environmentally sustainable production and use of energy in the transportation and manufacturing industry sectors. Imperative 3 of the Nuclear Energy R&D Roadmap, entitled “Enable a Transition Away from Fossil Fuels by Producing Process Heat for use in the Transportation and Industrial Sectors”, addresses this need. This document presents an Implementation Plan for R&D efforts related to this imperative. The expanded use of nuclear energy beyond the electrical grid will contribute significantly to overcoming the three inter-linked energy challenges facing U.S. industry: the rising and volatile prices for premium fossil fuels such as oil and natural gas, dependence on foreign sources for these fuels, and the risks of climate change resulting from carbon emissions. Nuclear energy could be used in the industrial and transportation sectors to: • Generate high temperature process heat and electricity to serve industrial needs including the production of chemical feedstocks for use in manufacturing premium fuels and fertilizer products, • Produce hydrogen for industrial processes and transportation fuels, and • Provide clean water for human consumption by desalination and promote wastewater treatment using low-grade nuclear heat as a useful additional benefit. Opening new avenues for nuclear energy will significantly enhance our nation’s energy

  9. Primary heat transport pump mechanical seal replacement strategy for Pickering B

    International Nuclear Information System (INIS)

    Chacinsi, V.

    1995-01-01

    Pickering Nuclear Generating Station is a CANDU PHWR eight unit station located on Lake Ontario. The station is divided into Pickering A (Units 1 to 4) and Pickering B (Units 5 to 8). Pickering B is the focus of this paper. Each unit is rated at 540 MWe. The Primary Heat Transport (PHT) system, which is used to cool the fuel, is divided into four quadrants. Each quadrant has four vertical Byron Jackson PHT main circulation pumps. Three pumps in each quadrant are required for normal operation, leaving one pump in each quadrant as a spare. Each Pickering PHT pump has a Byron Jackson Type SU two stage mechanical seal. The typical pressure breakdown across the seal is 8.7-4.5-1.0 MPa. Certain features of seal operation and the PHT system which influence seal replacement are discussed below. (author)

  10. ITER Generic Diagnostic Upper Port Plug Nuclear Heating and Personnel Dose Rate Assessment

    International Nuclear Information System (INIS)

    Feder, Russell E.; Youssef, Mahmoud Z.

    2009-01-01

    Neutronics analysis to find nuclear heating rates and personnel dose rates were conducted in support of the integration of diagnostics in to the ITER Upper Port Plugs. Simplified shielding models of the Visible-Infrared diagnostic and of a large aperture diagnostic were incorporated in to the ITER global CAD model. Results for these systems are representative of typical designs with maximum shielding and a small aperture (Vis-IR) and minimal shielding with a large aperture. The neutronics discrete-ordinates code ATTILA(reg s ign) and SEVERIAN(reg s ign) (the ATTILA parallel processing version) was used. Material properties and the 500 MW D-T volume source were taken from the ITER 'Brand Model' MCNP benchmark model. A biased quadrature set equivalent to Sn=32 and a scattering degree of Pn=3 were used along with a 46-neutron and 21-gamma FENDL energy subgrouping. Total nuclear heating (neutron plug gamma heating) in the upper port plugs ranged between 380 and 350 kW for the Vis-IR and Large Aperture cases. The Large Aperture model exhibited lower total heating but much higher peak volumetric heating on the upper port plug structure. Personnel dose rates are calculated in a three step process involving a neutron-only transport calculation, the generation of activation volume sources at pre-defined time steps and finally gamma transport analyses are run for selected time steps. ANSI-ANS 6.1.1 1977 Flux-to-Dose conversion factors were used. Dose rates were evaluated for 1 full year of 500 MW DT operation which is comprised of 3000 1800-second pulses. After one year the machine is shut down for maintenance and personnel are permitted to access the diagnostic interspace after 2-weeks if dose rates are below 100 (micro)Sv/hr. Dose rates in the Visible-IR diagnostic model after one day of shutdown were 130 (micro)Sv/hr but fell below the limit to 90 (micro)Sv/hr 2-weeks later. The Large Aperture style shielding model exhibited higher and more persistent dose rates. After 1

  11. Nuclear materials transportation workshops: USDOE outreach to local governments

    International Nuclear Information System (INIS)

    1987-01-01

    To provide direct outreach to local governments, the Transportation Management Division of the United States Department of Energy asked the Urban Consortium and its Energy Task Force to assemble representatives for two workshops focusing on the transport of nuclear materials. The first session, for jurisdictions east of the Mississippi River, was held in New Orleans on May 5--6, 1988; the second was conducted on June 6--7, 1988 in Denver for jurisdictions to the west. Twenty local government professionals with management or operational responsibility for hazardous materials transportation within their jurisdictions were selected to attend each workshop. The discussions identified five major areas of concern to local government professionals; coordination; training; information resources; marking and placarding; and responder resources. Integrated federal, state, and local levels of government emerged as a priority coordination issue along with the need for expanded availability of training and training resources for first-reponders

  12. Bibliographical survey of heat exchangers for nuclear power plants and problems of HTGR

    International Nuclear Information System (INIS)

    Yamao, Hiroyuki; Okamoto, Yoshizo; Sanokawa, Konomo

    1977-04-01

    The problems in development of heat exchangers for nuclear reactors have been examined in literature survey through Annual Index Subjects of NSA (Nuclear Science Abstracts) for the past ten years. R and D on heat exchangers for LMFBR, HTGR, LWR and HWR are on the increase. In the case of HTGRs, R and D on heat resisting materials including the corrosion and on hydrogen permeation of heat exchanger walls in high temperature pressure helium environment are important. Future R and D subjects for HTGR heat exchangers in showing the high temperature endurance are presented. (auth.)

  13. Numerical Modeling of the Vertical Heat Transport Through the Diffusive Layer of the Arctic Ocean

    Science.gov (United States)

    2013-03-01

    transport through thermohaline staircases in the Arctic region. Results revealed that vertical fluxes exceeded those of extant “four-thirds flux...vertical heat flux, thermohaline staircase 15. NUMBER OF PAGES 73 16. PRICE CODE 17. SECURITY CLASSIFICATION OF REPORT Unclassified 18...DNS) were conducted to assess the vertical heat transport through thermohaline staircases in the Arctic region. Results revealed that vertical

  14. Influence of tube and particle diameter on heat transport in packed beds

    NARCIS (Netherlands)

    Borkink, J.G.H.; Borkink, J.G.H.; Westerterp, K.R.

    1992-01-01

    Influence of the tube and particle diameter and shape, as well as their ratio, on the radial heat transport in packed beds has been studied. Heat transport experiments were performed with four different packings in three wall-cooled tubes, which differed in inner diameter only. Experimental values

  15. Transport phenomena in a sidewall-moving bottom-heated cavity ...

    Indian Academy of Sciences (India)

    The understanding of basic feature of energy transport from a heat source is important from the fundamental point of view as well as from various engineering and technological applications. To enrich the knowledge in this area, this paper presents energy transport phenomena from the heated bottom of an air-filled ...

  16. Enhanced heat transport in environmental systems using microencapsulated phase change materials

    Science.gov (United States)

    Colvin, D. P.; Mulligan, J. C.; Bryant, Y. G.

    1992-01-01

    A methodology for enhanced heat transport and storage that uses a new two-component fluid mixture consisting of a microencapsulated phase change material (microPCM) for enhanced latent heat transport is outlined. SBIR investigations for NASA, USAF, SDIO, and NSF since 1983 have demonstrated the ability of the two-component microPCM coolants to provide enhancements in heat transport up to 40 times over that of the carrier fluid alone, enhancements of 50 to 100 percent in the heat transfer coefficient, practically isothermal operation when the coolant flow is circulated in an optimal manner, and significant reductions in pump work.

  17. Status of non-electric nuclear heat applications: Technology and safety

    International Nuclear Information System (INIS)

    2000-11-01

    Nuclear energy plays an important role in electricity generation, producing 16% of the world's electricity at the beginning of 1999. It has proven to be safe, reliable, economical and has only a minimal impact on the environment. Most of the world's energy consumption, however, is in the form of heat. The market potential for nuclear heat was recognized early. Some of the first reactors were used for heat supply, e.g. Calder Hall (United Kingdom), Obninsk (Russian Federation), and Agesta (Sweden). Now, over 60 reactors are supplying heat for district heating, industrial processes and seawater desalination. But the nuclear option could be better deployed if it would provide a larger share of the heat market. In particular, seawater desalination using nuclear heat is of increasing interest to some IAEA Member States. In consideration of the growing experience being accumulated, the IAEA periodically reviews the progress and new developments in the field of nuclear heat applications. This publication summarizes the recent activities among Member States presented at a Technical Committee meeting in April 1999. The purpose of the meeting was to provide a forum for the exchange of up to date information on the prospect, design, safety and licensing aspects, and development of non-electrical applications of nuclear heat for industrial use. This mainly included seawater desalination and hydrogen production

  18. 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

  19. Strain dependence of the heat transport properties of graphene nanoribbons

    International Nuclear Information System (INIS)

    Emmeline Yeo, Pei Shan; Loh, Kian Ping; Gan, Chee Kwan

    2012-01-01

    Using a combination of accurate density-functional theory and a nonequilibrium Green’s function method, we calculate the ballistic thermal conductance characteristics of tensile-strained armchair (AGNR) and zigzag (ZGNR) edge graphene nanoribbons, with widths between 3 and 50 Å. The optimized lateral lattice constants for AGNRs of different widths display a three-family behavior when the ribbons are grouped according to N modulo 3, where N represents the number of carbon atoms across the width of the ribbon. Two lowest-frequency out-of-plane acoustic modes play a decisive role in increasing the thermal conductance of AGNR-N at low temperatures. At high temperatures the effect of tensile strain is to reduce the thermal conductance of AGNR-N and ZGNR-N. These results could be explained by the changes in force constants in the in-plane and out-of-plane directions with the application of strain. This fundamental atomistic understanding of the heat transport in graphene nanoribbons paves a way to effect changes in their thermal properties via strain at various temperatures. (paper)

  20. Experience gained in France on heat recovery from nuclear plant for agriculture and pisciculture

    International Nuclear Information System (INIS)

    Balligand, P.; Dumont, M.; Grauby, A.; Le Gouellec, P.

    1977-01-01

    For just five years the Commissariat a l'Energie Atomique has been interested in the use of thermal wastes from industrial installations particularly from nuclear power plants. Different types of pilot hothouses and their heating with water are presented in detail. The conclusions are that the thermal power plants owing to the Carnot principle release up to 60% of the thermal energy produced in the boiler into the environment but this thermal energy is at a very low temperature. In this paper it has been shown that agriculture and pisciculture can be satisfied with those low temperature waters. But transportation of this low temperature water is quite expensive and the total economy of a project has to be very carefully examined. (M.S.)

  1. Nuclear Technology Series. Course 4: Heat Transfer and Fluid Flow.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  2. Data module development for spent nuclear fuel transportation risk assessment

    International Nuclear Information System (INIS)

    Chen, S.Y.; Yuan, Y.C.; Zielen, A.J.

    1992-01-01

    In 1986, in compliance with the Nuclear Waste Policy Act of 1982, the U.S. Department of Energy (DOE) issued environmental assessments (EAs) of the potential repository sites for spent nuclear fuel (SNF) and high-level radioactive wastes. A major concern expressed in the public comments on the EAs was the need for a route-specific risk analysis. One approach to this concern was presented in 1987 by DOE's Office of Civilian Radioactive Waste Management (OCRWM) at a workshop on models for OCRWM transportation risk analysis in Salt Lake City, Utah. The DOE decided that analytical capabilities for a route-specific risk analysis should be maintained at the state level, thus requiring state-specific accident and agricultural output data. The DOE also decided that such an analysis warranted development of a more reliable data base for a number of key input parameters for transportation risk assessment. Such a data base would be compiled in a format acceptable for input into the existing transportation risk code RADTRAN. Seven data modules have been developed on the basis of these considerations. A system module was also developed for integrating the data modules and serves as a preprocessor of the RADTRAN input data file. The relationship of the modules to the RADTRAN input is shown

  3. Characterization of Heat-treated Clay Minerals in the Context of Nuclear Waste Disposal

    Science.gov (United States)

    Matteo, E. N.; Wang, Y.; Kruichak, J. N.; Mills, M. M.

    2015-12-01

    Clay minerals are likely candidates to aid in nuclear waste isolation due to their low permeability, favorable swelling properties, and high cation sorption capacities. Establishing the thermal limit for clay minerals in a nuclear waste repository is a potentially important component of repository design, as flexibility of the heat load within the repository can have a major impact on the selection of repository design. For example, the thermal limit plays a critical role in the time that waste packages would need to cool before being transferred to the repository. Understanding the chemical and physical changes, if any, that occur in clay minerals at various temperatures above the current thermal limit (of 100 °C) can enable decision-makers with information critical to evaluating the potential trade-offs of increasing the thermal limit within the repository. Most critical is gaining understanding of how varying thermal conditions in the repository will impact radionuclide sorption and transport in clay materials either as engineered barriers or as disposal media. A variety of repository-relevant clay minerals (illite, mixed layer illite/smectite, and montmorillonite), were heated for a range of temperatures between 100-1000 °C. These samples were characterized to determine surface area, mineralogical alteration, and cation exchange capacity (CEC). Our results show that for conditions up to 500 °C, no significant change occurs, so long as the clay mineral remains mineralogically intact. At temperatures above 500 °C, transformation of the layered silicates into silica phases leads to alteration that impacts important clay characteristics. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's Nation Nuclear Security Administration under contract DE-AC04-94AL85000. SAND Number: SAND2015-6524 A

  4. Transport of chemically bonded nuclear energy in a closed cycle with special consideration to energy disconnection

    International Nuclear Information System (INIS)

    Ossami, S.

    1976-01-01

    The article describes the utilisation of nuclear energy in the form of 'nuclear long-distance energy'. Heat produced by nuclear fission is bonded to a reversible chemical reaction (cracking gas) which release the heat again at the place of comsumption by catalytic transformation. The article deals in particular with the process of methane cracking/methanisation, the disconnection of the energy (heat) by the methanisation process and the decisive role of the methanisation catalyzers. (orig.) [de

  5. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

    1976-01-01

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  6. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dew Hey [Korea Institute of Nuclear and Safety, Taejon (Korea, Republic of); Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong [Chungnam National Univ., Taejon (Korea, Republic of)

    2002-03-15

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology.

  7. Analysis technology in the thick plate free drop impact, heat and thermal stress of the cask for radioactive material transport

    International Nuclear Information System (INIS)

    Lee, Dew Hey; Lee, Young Shin; Ryu, Chung Hyun; Kim, Hyun Su; Choi, Kyung Joo; Choi, Young Jin; Lee, Jae Hyung; Na, Jae Yun; Kim, Seong Jong

    2002-03-01

    In this study, The regulatory condition and analysis condition is analyzed for thick plate free drop, heat and thermal stress analysis to develop the safety assessment technology. Analysis is performed with finite element method which is one of the many analysis methods of the shipping cask. ANSYS, LS-DYNA3D and ABAQUS is suitable for thick plate free drop, heat and thermal stress analysis of the shipping cask. For the analysis model, the KSC-4 that is the shipping cask to transport spent nuclear fuel is investigated. The results of both LS-DYNA3D and ABAQUS for thick plate free drop and the results of ANSYS, LS-DYNA3D and ABAQUS for heat and thermal stress analysis is completely corresponded. And the integrity of the shipping cask is verified. Using this study, the reliable safety assessment technology is supplied to the staff. The efficient and reliable regulatory tasks is performed using the standard safety assessment technology

  8. Method and apparatus for nuclear heating of oil-bearing formations

    International Nuclear Information System (INIS)

    Alspaw, D.I.

    1979-01-01

    A method and apparatus are provided for using heat generated by absorption of radiation from nuclear waste materials to reduce the viscosity of petroleum products contained within a subsurface earth formation. The nuclear waste material is positioned in a salt water formation underlying the subsurface earth formation so that the radiation emitted by the material heats the salt water formation. conduction and convection transfer the heat to the subsurface earth formation, raising the temperature and thereby reducing the viscosity of the petroleum products. To prevent radioactive contamination within the salt water formation, the nuclear waste material may be encapsulated in a material selected to absorb alpha and beta radiation

  9. Pluto: Fluidized Transport of Tholins by Heating of the Subsurface

    Science.gov (United States)

    Cruikshank, Dale P.; Spohrer, Steven; Grundy, William M.; Moore, Jeffrey M.; Umurhan, Orkan M.; White, Oliver L.; Beyer, Ross A.; Dalle Ore, Cristina M.; Stern, S. A.; Young, Leslie; hide

    2017-01-01

    New Horizons images of Pluto show evidence of the transport of the colored non-ice component across the surface, with substantial accumulations in some areas of low elevation. The non-ice component is presumed to be tholin produced in the atmosphere as a precipitating aerosol, in the surface ices by photolysis or radiolysis, or both. We model the surface layer of N2 ice with varying amounts of incorporated tholin particles to explore the heating within the ice that occurs by the solid-state greenhouse effect. We find that in plausible models of the contaminated N2 surface ice the triple point temperature (63.15K) is reached at a depth of approximately less than 1m. At that depth the confining pressure of the ice column is much less than the triple point pressure (12.52 kPa), so N2 should convert to the gas phase, exerting pressure on the overburden. When the gas pressure exceeds the strength of the confining ice, a breakout on the surface will occur, fluidizing fragments of ice and its contaminants that are then free to flow downhill, rafted on entrained gas, similar in some ways to the pyroclastic volcanic phenomenon known as nuée ardente. The digital elevation map of Pluto made from stereo images shows some surface regions that may have been stripped of the N2 layer, exposing H2O ice (presumed to be bedrock) below, with a corresponding accumulation of dark material that was that was the previously entrained particulate tholin. Accumulations of tholin are found associated with some of the fossae, and some cover preexisting topography to depths of up to a few hundred meters.

  10. Dynamics of miRNA biogenesis and nuclear transport

    OpenAIRE

    Kotipalli, Aneesh; Gutti, Ravikumar; Mitra, Chanchal K.

    2016-01-01

    MicroRNAs (miRNAs) are short noncoding RNA sequences ~22 nucleotides in length that play an important role in gene regulation-transcription and translation. The processing of these miRNAs takes place in both the nucleus and the cytoplasm while the final maturation occurs in the cytoplasm. Some mature miRNAs with nuclear localisation signals (NLS) are transported back to the nucleus and some remain in the cytoplasm. The functional roles of these miRNAs are seen in both the nucleus and the cyto...

  11. Policy issues of transporting spent nuclear fuel by rail

    International Nuclear Information System (INIS)

    Spraggins, H.B.

    1994-01-01

    The topic of this paper is safe and economical transportation of spent nuclear fuel by rail. The cost of safe movement given the liability consequences in the event of a rail accident involving such material is the core issue. Underlying this issue is the ability to access the risk probability of such an accident. The paper delineates how the rail industry and certain governmental agencies perceive and assess such important operational, safety, and economic issues. It also covers benefits and drawbacks of dedicated and regular train movement of such materials

  12. Molten Chloride Salts for Heat Transfer in Nuclear Systems

    Science.gov (United States)

    Ambrosek, James Wallace

    2011-12-01

    A forced convection loop was designed and constructed to examine the thermal-hydraulic performance of molten KCl-MgCl2 (68-32 at %) salt for use in nuclear co-generation facilities. As part of this research, methods for prediction of the thermo-physical properties of salt mixtures for selection of the coolant salt were studied. In addition, corrosion studies of 10 different alloys were exposed to the KCl-MgCl2 to determine a suitable construction material for the loop. Using experimental data found in literature for unary and binary salt systems, models were found, or developed to extrapolate the available experimental data to unstudied salt systems. These property models were then used to investigate the thermo-physical properties of the LINO3-NaNO3-KNO 3-Ca(NO3), system used in solar energy applications. Using these models, the density, viscosity, adiabatic compressibility, thermal conductivity, heat capacity, and melting temperatures of higher order systems can be approximated. These models may be applied to other molten salt systems. Coupons of 10 different alloys were exposed to the chloride salt for 100 hours at 850°C was undertaken to help determine with which alloy to construct the loop. Of the alloys exposed, Haynes 230 had the least amount of weight loss per area. Nickel and Hastelloy N performed best based on maximum depth of attack. Inconel 625 and 718 had a nearly uniform depletion of Cr from the surface of the sample. All other alloys tested had depletion of Cr along the grain boundaries. The Nb in Inconel 625 and 718 changed the way the Cr is depleted in these alloys. Grain-boundary engineering (GBE) of Incoloy 800H improved the corrosion resistance (weight loss and maximum depth of attack) by nearly 50% as compared to the as-received Incoloy 800H sample. A high temperature pump, thermal flow meter, and pressure differential device was designed, constructed and tested for use in the loop, The heat transfer of the molten chloride salt was found to

  13. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  14. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    International Nuclear Information System (INIS)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-01-01

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in a circular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mass velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel

  15. 25 CFR 170.900 - What is the purpose of the provisions relating to transportation of hazardous and nuclear waste?

    Science.gov (United States)

    2010-04-01

    ... transportation of hazardous and nuclear waste? 170.900 Section 170.900 Indians BUREAU OF INDIAN AFFAIRS... and Nuclear Waste Transportation § 170.900 What is the purpose of the provisions relating to transportation of hazardous and nuclear waste? Sections 170.900 through 170.907 on transportation of nuclear and...

  16. Encapsulated nuclear heat source reactors for energy security

    International Nuclear Information System (INIS)

    Greenspan, E.; Susplugas, A.; Hong, S.G.; Monti, L.; Sumini, M.; Okawa, T.

    2006-01-01

    A spectrum of Encapsulated Nuclear Heat Source (ENHS) reactors have been conceptually designed over the last few years; they span a power range from 10 MWe to -200 MWe and consider a number of coolants and fuel types. Common features of all these designs include very long life cores - exceeding 20 effective full power years; nearly zero burnup reactivity swing; natural circulation; superb safety; autonomous load following capability; simplicity of operation and maintenance. ENHS reactors could be of particular interest for providing electricity, thermal energy and, possibly, desalinated water to communities that are not connected to a central electricity grid such as to many pacific islands and to remote communities in the mainland of different countries. ENHS reactors provide energy security by virtue of a couple of features: (1) Once an ENHS reactor is commissioned, the community has assured clean energy supply for at least 20 years without needing fuel supply. (2) The energy value of the fuel loaded (in the factory) in the ENHS module is preserved; what is needed for generating energy for additional 20+ years is to remove the fission products, add depleted uranium for makeup fuel, refabricate fuel rods and load into a new module. This fuel recycling is envisioned done by either the supplier country or by a regional or international fuel cycle centre. As the ENHS module is replaced at its entirety at the end of the core life - that is brought about by radiation damage, the ENHS plant life is likely to last for over 100 years. The above features also offer exceptional stability in the price of energy generated by the ENHS reactor. The reference ENHS design will be described followed by a brief description of the design options developed and a summary of their performance characteristics

  17. Experimental investigation on heat transport in gravel-sand materials

    DEFF Research Database (Denmark)

    Maureschat, Gerald; Heller, Alfred

    1997-01-01

    in sand-gravel material, the storage media is to be water satured. In this case, handling of such material on site is rather complex. The conduction is highly dependent on the thermal properties of the storage media and so is the overall thermal performance of a storage applying such media. For sandy...... out in a small size experiment. The experiment consists of a highly insulated box filled with two kinds of sand material crossed by a plastic heat pipe. Heat transfer is measured under dry and water satured conditions in a cross-section.The conclusions are clear. To obtain necessary heat conduction......The project is a basic study on the expected thermal behaviour of gravel storage initiated as a part of a research and demonstration gravel storage for seasonal heat storage.The goal of the investigation is to determine the heat transfer between heat pipes and sand-gravel storage media by carrying...

  18. Meeting Czechoslovak demands for heat in long-term prospective, especially with regard to nuclear sources

    International Nuclear Information System (INIS)

    Klail, M.

    1988-01-01

    The development was studied of heat demand in the CSSR till the year 2030. The ratio of centralized and decentralized heat supply is currently 60 to 40; in the future a slight increase is expected in the decentralized type of heat supply, mainly as a result of more intensive use of natural gas. In 2030, 710 PU of centralized heat should be produced. A decisive element in meeting the demand will be a growing proportion of combined production of electric power and heat by nuclear power plants. The installed capacity of the nuclear power plants in 2030 should range between 23 and 41 thousand MW, the production of electric power in these plants should be 193 to 238 TWh/y. 109 territorial areas potentially suitable for use of heat from nuclear sources were selected. They were included in 19 regions of which 9 should in the year 2010 be linked to heat supply from nuclear power plants that will be in operation. It is expected that in the year 2030, nuclear sources will supply 250 PU of centralized heat. (Z.M.). 2 tabs., 14 refs

  19. Reprint of "Nuclear transport factors: global regulation of mitosis".

    Science.gov (United States)

    Forbes, Douglass J; Travesa, Anna; Nord, Matthew S; Bernis, Cyril

    2015-06-01

    The unexpected repurposing of nuclear transport proteins from their function in interphase to an equally vital and very different set of functions in mitosis was very surprising. The multi-talented cast when first revealed included the import receptors, importin alpha and beta, the small regulatory GTPase RanGTP, and a subset of nuclear pore proteins. In this review, we report that recent years have revealed new discoveries in each area of this expanding story in vertebrates: (a) The cast of nuclear import receptors playing a role in mitotic spindle regulation has expanded: both transportin, a nuclear import receptor, and Crm1/Xpo1, an export receptor, are involved in different aspects of spindle assembly. Importin beta and transportin also regulate nuclear envelope and pore assembly. (b) The role of nucleoporins has grown to include recruiting the key microtubule nucleator – the γ-TuRC complex – and the exportin Crm1 to the mitotic kinetochores of humans. Together they nucleate microtubule formation from the kinetochores toward the centrosomes. (c) New research finds that the original importin beta/RanGTP team have been further co-opted by evolution to help regulate other cellular and organismal activities, ranging from the actual positioning of the spindle within the cell perimeter, to regulation of a newly discovered spindle microtubule branching activity, to regulation of the interaction of microtubule structures with specific actin structures. (d) Lastly, because of the multitudinous roles of karyopherins throughout the cell cycle, a recent large push toward testing their potential as chemotherapeutic targets has begun to yield burgeoning progress in the clinic. Copyright © 2015 Elsevier Ltd. All rights reserved.

  20. Ballistic near-field heat transport in dense many-body systems

    Science.gov (United States)

    Latella, Ivan; Biehs, Svend-Age; Messina, Riccardo; Rodriguez, Alejandro W.; Ben-Abdallah, Philippe

    2018-01-01

    Radiative heat transport mediated by near-field interactions is known to be superdiffusive in dilute, many-body systems. Here we use a generalized Landauer theory of radiative heat transfer in many-body planar systems to demonstrate a nonmonotonic transition from superdiffusive to ballistic transport in dense systems. We show that such a transition is associated to a change of the polarization of dominant modes. Our findings are complemented by a quantitative study of the relaxation dynamics of the system in the different regimes of heat transport. This result could have important consequences on thermal management at nanoscale of many-body systems.

  1. Nonlinear heat and particle transport due to collisional drift waves

    Energy Technology Data Exchange (ETDEWEB)

    Nishi-kawa, K.I.; Hatori, T.; Terashima, Y.

    1977-03-01

    The nonlinear evolution of unstable modes which govern transport processes in magnetically confined plasmas were investigated. A nonlinear theory of unstable collisional drift wave, and the consequent nonlinear transport were extended to include electron and ion temperature gradients. Thermal transport properties are discussed and basic equations are given.

  2. ANALYZING NUMERICAL ERRORS IN DOMAIN HEAT TRANSPORT MODELS USING THE CVBEM.

    Science.gov (United States)

    Hromadka, T.V.

    1987-01-01

    Besides providing an exact solution for steady-state heat conduction processes (Laplace-Poisson equations), the CVBEM (complex variable boundary element method) can be used for the numerical error analysis of domain model solutions. For problems where soil-water phase change latent heat effects dominate the thermal regime, heat transport can be approximately modeled as a time-stepped steady-state condition in the thawed and frozen regions, respectively. The CVBEM provides an exact solution of the two-dimensional steady-state heat transport problem, and also provides the error in matching the prescribed boundary conditions by the development of a modeling error distribution or an approximate boundary generation.

  3. MODELING OF THE GROUNDWATER TRANSPORT AROUND A DEEP BOREHOLE NUCLEAR WASTE REPOSITORY

    Energy Technology Data Exchange (ETDEWEB)

    N. Lubchenko; M. Rodríguez-Buño; E.A. Bates; R. Podgorney; E. Baglietto; J. Buongiorno; M.J. Driscoll

    2015-04-01

    The concept of disposal of high-level nuclear waste in deep boreholes drilled into crystalline bedrock is gaining renewed interest and consideration as a viable mined repository alternative. A large amount of work on conceptual borehole design and preliminary performance assessment has been performed by researchers at MIT, Sandia National Laboratories, SKB (Sweden), and others. Much of this work relied on analytical derivations or, in a few cases, on weakly coupled models of heat, water, and radionuclide transport in the rock. Detailed numerical models are necessary to account for the large heterogeneity of properties (e.g., permeability and salinity vs. depth, diffusion coefficients, etc.) that would be observed at potential borehole disposal sites. A derivation of the FALCON code (Fracturing And Liquid CONvection) was used for the thermal-hydrologic modeling. This code solves the transport equations in porous media in a fully coupled way. The application leverages the flexibility and strengths of the MOOSE framework, developed by Idaho National Laboratory. The current version simulates heat, fluid, and chemical species transport in a fully coupled way allowing the rigorous evaluation of candidate repository site performance. This paper mostly focuses on the modeling of a deep borehole repository under realistic conditions, including modeling of a finite array of boreholes surrounded by undisturbed rock. The decay heat generated by the canisters diffuses into the host rock. Water heating can potentially lead to convection on the scale of thousands of years after the emplacement of the fuel. This convection is tightly coupled to the transport of the dissolved salt, which can suppress convection and reduce the release of the radioactive materials to the aquifer. The purpose of this work has been to evaluate the importance of the borehole array spacing and find the conditions under which convective transport can be ruled out as a radionuclide transport mechanism

  4. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    Arnish, J.J.; Monette, F.; LePoire, D.; Biwer, B.M.

    1996-01-01

    The RISKIND software package is used to estimate radiological consequences of severe accident scenarios involving the transportation of spent nuclear fuel. Radiological risks are estimated for both a collective population and a maximally exposed individual based on representative truck and rail cask designs described in the U.S. Nuclear Regulatory Commission (NRC) modal study. The estimate of collective population risk considers all possible environmental pathways, including acute and long-term exposures, and is presented in terms of the 50-y committed effective dose equivalent. Radiological risks to a maximally exposed individual from acute exposure are estimated and presented in terms of the first year and 50-y committed effective dose equivalent. Consequences are estimated for accidents occurring in rural and urban population areas. The modeled pathways include inhalation during initial passing of the radioactive cloud, external exposure from a reduction of the cask shielding, long-term external exposure. from ground deposition, and ingestion from contaminated food (rural only). The major pathways and contributing radionuclides are identified, and the effects of possible mitigative actions are discussed. The cask accident responses and the radionuclide release fractions are modeled as described in the NRC modal study. Estimates of severe accident probabilities are presented for both truck and rail modes of transport. The assumptions made in this study tend to be conservative; however, a set of multiplicative factors are identified that can be applied to estimate more realistic conditions

  5. Measuring the linear heat generation rate of a nuclear reactor fuel pin

    International Nuclear Information System (INIS)

    Smith, R.D.

    1981-01-01

    A miniature gamma thermometer is described which is capable of travelling through bores distributed in an array through a nuclear reactor core and measure the linear heat generation rate of the fuel pins. (U.K.)

  6. Heat balance in the catalyst layer and the boundary condition for heat transport equation in a low-temperature fuel cell

    Science.gov (United States)

    Kulikovsky, A. A.

    We consider heat balance in the catalyst layer of a low-temperature fuel cell. Based on the heat transport equation in the layer, the exact boundary condition for the problem of heat transport in the cell is obtained. The limits of validity of the resulting expression are discussed.

  7. Performances of nuclear power plants for combined production of electricity and hot water for district heating

    International Nuclear Information System (INIS)

    Bronzen, S.

    The possibilities for using nuclear power plants for combined production of heat and power seem to be very good in the future. With the chosen 600 MWsub (e) BWR plant a heat output up to 1200 MW can be arranged. An alternative, consisting of steam extractions from the low-pressure turbine, offers a flexible solution for heat and power generation. With this alternative the combined plant can use components from normal condensing nuclear power plants. The flexible extraction design also offers a real possibility for using the combined plant in electric peak generation. However, urban siting requires long distance heat transmission and the pipe design for this transmission is a major problem when planning and optimizing the whole nuclear combined heat and power plant. (author)

  8. Design aspects of LTE desalination plant utilizing waste heat from nuclear research reactor

    International Nuclear Information System (INIS)

    Rao, I.S.; Srivastava, V.K.; Tewari, P.K.

    2004-01-01

    Nuclear research reactors produce significant amount of low quality waste heat which can be utilized for producing high quality water from seawater by coupling a Low Temperature Evaporation (LTE) desalination unit. Salient features and design considerations of the desalination plant coupled to the nuclear research reactor and the performance of the desalination plant under varying operational conditions applicable to waste heat utilization from the reactor are discussed. Chemical and radioactive analysis of the product water is given to indicate the usefulness of the water to meet the demineralized water makeup requirements of the reactor. The general scheme of integrating desalination plant with the nuclear research reactor is also presented. This LTE desalination plant utilizing waste heat from a nuclear research reactor is the first of its kind and is a demonstration of safety and economics of nuclear desalination technology as a viable alternative to produce demineralised water from seawater. (author)

  9. Nuclear thermal propulsion transportation systems for lunar/Mars exploration

    International Nuclear Information System (INIS)

    Clark, J.S.; Borowski, S.K.; Mcilwain, M.C.; Pellaccio, D.G.

    1992-09-01

    Nuclear thermal propulsion technology development is underway at NASA and DoE for Space Exploration Initiative (SEI) missions to Mars, with initial near-earth flights to validate flight readiness. Several reactor concepts are being considered for these missions, and important selection criteria will be evaluated before final selection of a system. These criteria include: safety and reliability, technical risk, cost, and performance, in that order. Of the concepts evaluated to date, the Nuclear Engine for Rocket Vehicle Applications (NERVA) derivative (NDR) is the only concept that has demonstrated full power, life, and performance in actual reactor tests. Other concepts will require significant design work and must demonstrate proof-of-concept. Technical risk, and hence, development cost should therefore be lowest for the concept, and the NDR concept is currently being considered for the initial SEI missions. As lighter weight, higher performance systems are developed and validated, including appropriate safety and astronaut-rating requirements, they will be considered to support future SEI application. A space transportation system using a modular nuclear thermal rocket (NTR) system for lunar and Mars missions is expected to result in significant life cycle cost savings. Finally, several key issues remain for NTR's, including public acceptance and operational issues. Nonetheless, NTR's are believed to be the next generation of space propulsion systems - the key to space exploration

  10. Application and performance requirements of hydraulic expanding technique for nuclear heat-exchanger

    International Nuclear Information System (INIS)

    Zhang Zhenhua

    2013-01-01

    The paper introduces the process and advantages of the hydraulic expanding method, and analyzes the mechanism and method of hydraulic expansion by graphical methods. Meanwhile, three performance requirements of the hydraulic expanding are described. Especially through the introduction to a typical nuclear heat exchanger tube expanding process, the importance of hydraulic expansion of tubes in design and manufacturing of nuclear heat exchanger is further illustrated. (authors)

  11. WHIST transport analysis of high neutron production, ICRH heated, pellet fueled jet plasmas

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Milora, S.L.; Tolliver, J.S.; Phillips, C.K.

    1988-01-01

    The WHIST 1-1/2-D predictive transport code is used to model the particle and energy transport of JET pellet-fueled, ICRH-heated plasmas. Pellet injection during the current rise phase was used to produce strong central peaking of the particle density followed by central ICRH heating and led to transient period of enhanced confinement. The evolution of the density profile as well as the electron and ion temperature profiles and strong ICRH heating conditions are examined during this period of enhanced confinement in the context of models for particle and energy transport. Because WHIST is a predictive transport code, it requires models for particle and energy sources and transport coefficients. The analysis procedure thus consists of modeling the particle source terms (pellets, gas, and recycled neutrals), energy source terms (ohmic and ICRH heating), and energy loss terms (primarily radiation), and varying the transport models until the best qualitative and quantitative agreement is obtained between calculated and observed quantities. We find that plasma behavior is well described during the first second of ICRH heating following pellet injection by the same transport coefficients that describe the ohmic plasma. The distinction between electron and ion thermal losses depends on the relative heating rates of electrons and ions as determined by the ICRH model, as well as the radiation losses. 10 refs., 4 figs

  12. Changes in Tropical Precipitation at the Mid-Holocene: Role of the Oceanic Heat Transport

    Science.gov (United States)

    Liu, X.; Battisti, D. S.; Donohoe, A.

    2015-12-01

    There is ample geological and geochemical evidence that precipitation in the tropics is largely different from today at the mid-Holocene, an era roughly 6,000 years ago when the Northern Hemisphere summer (winter) insolation was stronger (weaker) than today. These insolation differences are caused mainly by the precession of the earth's rotational axis, or called "precessional forcing". Using the mid-Holocene experiments of PMIP3, we studied changes in the zonal mean tropical precipitation, and its associated change in cross-equatorial energy transport. A northward movement of the zonal mean precipitation in the mid-Holocene is seen in 10 out of 13 PMIP3 models, with a correspondingly anomalous southward atmospheric heat transport across the equator. The slope is 3.0º per PW, close to the estimate given by Donohoe et al. (2013). The changes in cross-equatorial atmospheric heat transport are dictated by changes in the hemispheric asymmetry of heating from the surface, which in turn are associated with changes in the cross-equatorial oceanic heat transport: an anomalous northward oceanic heat transport at the equator is seen in all of the PMIP3 models. Analysis on this anomalous oceanic heat transport reveals that changes in the wind-driven gyre in the Pacific Ocean are primarily responsible for the changes in cross-equatorial ocean heat transport. Specifically, stronger easterly anomalies north of the equator in the western Pacific drives an anomalous northward mass transport, and therefore accomplishes an anomalous northward heat transport across the equator by acting on the asymmetric mean-state zonal temperature. The wind anomalies responsible for this anomalous ocean heat transport are seen in every PMIP3 model, as well as an ECHAM4-slab ocean model, indicating that it is atmospherically driven and independent of the changes in ocean heat transport. It also explains the consistency of ocean heat transport change, and eventually the relative consistency of zonal

  13. 75 FR 45608 - Blue Ribbon Commission on America's Nuclear Future, Transportation and Storage Subcommittee

    Science.gov (United States)

    2010-08-03

    ..., processing, and disposal of civilian and defense spent nuclear fuel and nuclear waste. The Co-chairs of the... the way in which it is storing used nuclear fuel and high level waste while one or more final disposal... Ribbon Commission on America's Nuclear Future, Transportation and Storage Subcommittee AGENCY: Department...

  14. District heating

    International Nuclear Information System (INIS)

    1989-03-01

    The papers presented at this meeting dealt with an international comparison of district heating, the Swiss district heating network, political aspects of nuclear district heating, nuclear and non-nuclear sources for district heating. 17 figs., 6 tabs

  15. Heat conduction through geological mattresses from cells storing mean activity and long life nuclear wastes

    International Nuclear Information System (INIS)

    Lajoie, D.; Raffourt, C.; Wendling, J.

    2010-01-01

    Document available in extended abstract form only. ANDRA ordered in 2008 a campaign of numerical simulations to assess the efficiency of the ventilation system designed for cells storing mean activity and long life nuclear wastes. Numerical models were performed by ACRIIN as research engineering office. The main objectives were to assess the risks of atmospheric explosions due to high rate of hydrogen and to determine the efficiency of the system to evacuate released heat from storage packages. Further calculations have been carried out to evaluate temperature gradients in the surrounding geological medium. Three-dimensional numerical models of a reference cell were built to simulate the air flow injected at the cell entrance and retrieved and the other extremity. The reference case is based on a cell full of storage packages, with rows and columns of packages methodically ordered. Analytic and numerical calculations have been performed introducing progressively each complex physical phenomenon in order to dissociate origins of transport of released mass or heat. Three kinds of flows have been physically distinguished: 1) Ventilation in a cell with storage package that are thermally inert, i.e. no heat release, but with hydrogen release. 2) Flow in a cell with storage packages that emit heat and warm the injected air, supposing that no heat were lost towards the surrounding concrete walls of the cell. 3) Air Flow warmed by the storage packages with heat losses towards concrete walls and geological medium. Simulations with absence of thermal effects allowed the knowledge of main topics of the ventilation air flows that may be synthesized as follows: - Flows infiltrate clearances between piles and rows of storage packages. Such apertures are a few centimetres wide. The flow is disorganised between the first rows, with distribution in both transversal and longitudinal directions. After a few tens of rows, the flow reaches its hydraulic equilibrium, with a nearly pure

  16. Glas generator for the steam gasification of coal with nuclear generated heat

    International Nuclear Information System (INIS)

    Buchner, G.

    1980-01-01

    The use of heat from a High Temperature Reactor (HTR) for the steam gasification of coal saves coal, which otherwise is burnt to generate the necessary reaction heat. The gas generator for this process, a horizontal pressure vessel, contains a fluidized bed of coal and steam. An ''immersion-heater'' type of heat exchanger introduces the nuclear generated heat to the process. Some special design problems of this gasifier are presented. Reference is made to the present state of development of the steam gasification process with heat from high temperature reactors. (author)

  17. Heat transport in the XXZ spin chain: from ballistic to diffusive regimes and dephasing enhancement

    International Nuclear Information System (INIS)

    Mendoza-Arenas, J J; Al-Assam, S; Clark, S R; Jaksch, D

    2013-01-01

    In this work we study the heat transport in an XXZ spin-1/2 Heisenberg chain with homogeneous magnetic field, incoherently driven out of equilibrium by reservoirs at the boundaries. We focus on the effect of bulk dephasing (energy-dissipative) processes in different parameter regimes of the system. The non-equilibrium steady state of the chain is obtained by simulating its evolution under the corresponding Lindblad master equation, using the time evolving block decimation method. In the absence of dephasing, the heat transport is ballistic for weak interactions, while being diffusive in the strongly interacting regime, as evidenced by the heat current scaling with the system size. When bulk dephasing takes place in the system, diffusive transport is induced in the weakly interacting regime, with the heat current monotonically decreasing with the dephasing rate. In contrast, in the strongly interacting regime, the heat current can be significantly enhanced by dephasing for systems of small size. (paper)

  18. Subcooled He II heat transport in the channel with abrupt contractions/enlargements

    International Nuclear Information System (INIS)

    Maekawa, R.; Iwamoto, A.; Hamaguchi, S.; Mito, T.

    2002-01-01

    Heat transport mechanisms for subcooled He II in the channel with abrupt contractions and/or enlargements have been investigated under steady state conditions. The channel, made of G-10, contains various contraction geometries to simulate the cooling channel of a superconducting magnet. In other words, contractions are periodically placed along the channel to simulate the spacers within the magnet winding. A copper block heater inputs the heat to the channel from one end, while the other end is open to the He II bath. Temperature profiles were measured with temperature sensors embedded in the channel as a function of heat input. Calculations were performed using a simple one-dimensional turbulent heat transport equation and with geometric factor consideration. The effects on heat transport mechanisms in He II caused by abrupt change of channel geometry and size are discussed

  19. Safeguards systems concepts for nuclear material transportation. Final report

    International Nuclear Information System (INIS)

    Baldonado, O.C.; Kevany, M.; Rodney, D.; Pitts, D.; Mazur, M.

    1977-09-01

    The report describes the development of system concepts for the safeguarding of special strategic nuclear materials (SNM) against malevolent adversary action during the interfacility transport of the SNM. The methodology used includes techniques for defining, classifying, and analyzing adversary action sequences; defining safeguards system components; assessing the vulnerability of various safeguards systems and their component parts to the potential adversary action sequences, and conceptualizing system design requirements. The method of analysis is based primarily on a comparison of adversary actions with safeguards measures, to estimate vulnerability. Because of the paucity of the data available for assessing vulnerability, the Delphi approach was used to generate data: values were estimated in a structured exercise by a panel of experts in the safeguards and terrorist fields. It is concluded that the probability of successful attack against a truck/escort convoy manned by well-trained, well-armed personnel is low enough to discourage all but the strongest adversaries. Secrecy of operations and careful screening of personnel are very important. No reliance should be placed on current capabilities of local law enforcement agencies. The recommendation of the study is the use of road transport in the near future and air transport at a later time when the number of shipments reaches a level to justify it, and when present safety problems are resolved

  20. FEFLOW finite element modeling of flow, mass and heat transport in porous and fractured media

    CERN Document Server

    Diersch, Hans-Jörg G

    2013-01-01

    Placing advanced theoretical and numerical methods in the hands of modeling practitioners and scientists, this book explores the FEFLOW system for solving flow, mass and heat transport processes in porous and fractured media. Offers applications and exercises.

  1. Valorization of the energy potential of fossil and fissile fuels for heat production: dual-purpose power plants and heat-producing nuclear reactors

    International Nuclear Information System (INIS)

    Lavite, Michel.

    1975-07-01

    The heat market is analyzed briefly within the French context: present structures and characteristics of the market, current means of heat production, predictable trend of the demand. The possible applications of nuclear energy to heat production, through the agency of combined electricity-steam stations or heat-producing stations, are then examined. Nuclear solutions are compared with others from the technico-economic and ecological wiewpoints and an estimate fo their respective impacts on the energy balance is attempted [fr

  2. Apparatus to simulate nuclear heating in advanced fuels

    International Nuclear Information System (INIS)

    Wrona, B.J.; Galvin, T.M.; Johanson, E.

    1976-10-01

    A direct-electrical-heating apparatus has been built to simulate in-reactor temperature gradients and heating conditions in both the mixed nitrides and carbides of uranium and plutonium. The apparatus has the capability for the investigation and direct observation of fuel-behavior phenomena that should significantly enlarge the data base on mixed carbides and nitrides at temperatures near and above their melting points. In addition to heating UC, results of prooftests showed that the apparatus has the capability to heat graphite, 30 vol % ZrC in graphite, B 4 C control-rod pellets, and stainless steel

  3. Calorimeter measures high nuclear heating rates and their gradients across a reactor test hole

    Science.gov (United States)

    Burwell, D.; Coombe, J. R.; Mc Bride, J.

    1970-01-01

    Pedestal-type calorimeter measures gamma-ray heating rates from 0.5 to 7.0 watts per gram of aluminum. Nuclear heating rate is a function of cylinder temperature change, measured by four chromel-alumel thermocouples attached to the calorimeter, and known thermoconductivity of the tested material.

  4. Nuclear heat applications: design aspects and operating experience. Proceedings of four technical meetings held between December 1995 and April 1998

    International Nuclear Information System (INIS)

    1998-11-01

    Proven to be safe, reliable, economic and having minimum impact on the environment, nuclear energy is playing an important role in electricity generation producing 175 of the world's electricity. But since most of the world's energy consumption is in the form of heat the market for nuclear heat has already been recognised. Considering the growing experience in application of power reactors for district heating, industrial processes and water desalination IAEA is periodically reviewing progress and new developments of nuclear heat applications. This proceedings includes the papers presented at the following four meetings: Advisory group meeting and consultancy on experience with nuclear heat applications: district heating, process heat and desalination, 13-15 December 1995 and 7-9 february 1996; Advisory group meeting on technology, design and safety aspects of non-electrical application of nuclear energy, 20-24 October 1997; Advisory group meeting on operational modes of nuclear desalination plants, 3-5 November 1997; Advisory group meeting on materials and equipment for the coupling interfaces of nuclear reactors with desalination and district heating plants, 21-23 April 1998. It is structured according to the subject areas: (1) design an safety aspects of nuclear heat application, (2) operational and material aspects of nuclear heat application and (3) operational experience with nuclear heat application. Each paper is described by a separate abstract

  5. New Parametrization for Heat Transport Through Diffusive Convection Interface

    Science.gov (United States)

    Guo, Shuang-Xi; Cen, Xian-Rong; Zhou, Sheng-Qi

    2018-02-01

    A laboratory experiment of two-layer diffusive convection (DC) within a rectangular cell is reported. The run-down evolution of two mixed layers and their interface is exhibited by the temperature and salinity structures, where the interface thicknesses of temperature and salinity increase with time as hT˜t0.42 and hS˜t0.43. Heat flux across the interface is respectively evaluated in terms of the heat energy conservation and the assumption of conductive interface. Both of them give the consistent heat flux values. When the 4/3 scaling between heat flux F and temperature difference across the interface ΔT holds (F˜C>(Rρ>)ΔT4/3, where C>(Rρ>) is function of density ratio Rρ), previous heat flux parameterizations are found to have low coefficient of determination (r2≤0.5) with our heat flux data within 1.6law function of Rρ, and the best fitting is C>(Rρ>)=0.081>(Rρ-1>)-1.28, resulting in r2 as high as 0.74. By using all available heat flux data over a wide Rρ range (1.2-27.6), including those collected in previous literatures, C>(Rρ>) is revised as C>(Rρ>)=0.065>(Rρ-1>)-1.20 accompanied by r2=0.70 and a reduced chi-square χ2=0.91. Alternatively, C>(Rρ>) can be derived from previous DC layer thickness parameterization, which is expressed as C>(Rρ>)=0.076>(Rρ-1>)-4/3. This formula is also superior to previous parameterizations in the evaluation of heat flux, indicated by high r2 and low χ2.

  6. Magnetically Modulated Heat Transport in a Global Simulation of Solar Magneto-convection

    Energy Technology Data Exchange (ETDEWEB)

    Cossette, Jean-Francois [Laboratory for Atmospheric and Space Physics, Campus Box 600, University of Colorado, Boulder, CO 80303 (United States); Charbonneau, Paul [Département de Physique, Université de Montréal, C.P. 6128, Succ. Centre-Ville, Montréal, QC H3C 3J7 (Canada); Smolarkiewicz, Piotr K. [European Centre for Medium-Range Weather Forecasts, Reading, RG2 9AX (United Kingdom); Rast, Mark P., E-mail: Jean-Francois.Cossette@lasp.colorado.edu, E-mail: paulchar@astro.umontreal.ca, E-mail: smolar@ecmwf.int, E-mail: Mark.Rast@lasp.colorado.edu [Department of Astrophysical and Planetary Sciences, Laboratory for Atmospheric and Space Physics, Campus Box 391, University of Colorado, Boulder, CO 80303 (United States)

    2017-05-20

    We present results from a global MHD simulation of solar convection in which the heat transported by convective flows varies in-phase with the total magnetic energy. The purely random initial magnetic field specified in this experiment develops into a well-organized large-scale antisymmetric component undergoing hemispherically synchronized polarity reversals on a 40 year period. A key feature of the simulation is the use of a Newtonian cooling term in the entropy equation to maintain a convectively unstable stratification and drive convection, as opposed to the specification of heating and cooling terms at the bottom and top boundaries. When taken together, the solar-like magnetic cycle and the convective heat flux signature suggest that a cyclic modulation of the large-scale heat-carrying convective flows could be operating inside the real Sun. We carry out an analysis of the entropy and momentum equations to uncover the physical mechanism responsible for the enhanced heat transport. The analysis suggests that the modulation is caused by a magnetic tension imbalance inside upflows and downflows, which perturbs their respective contributions to heat transport in such a way as to enhance the total convective heat flux at cycle maximum. Potential consequences of the heat transport modulation for solar irradiance variability are briefly discussed.

  7. Sensitivity analysis for heat diffusion in a fin on a nuclear fuel element

    International Nuclear Information System (INIS)

    Tito, Max Werner de Carvalho

    2001-11-01

    projected to gas-cooled nuclear reactors to compensate the low coolant thermal transport efficiency. The model is described by the temperature distribution equation and the further specific boundary conditions. The adjoint system is used to determine the sensitivity coefficients to the case of interest. Both, the direct model and the perturbative formalism resultant equations are solved. The heat flow rate on a point of the fin and the average temperature excess were the response functionals studied. The half thickness, the thermal conductivity and heat transfer coefficients and the heat flow from the base material were the parameters of interest to the sensitivity analysis. The results obtained through the perturbative method and the direct variation presented, in a general form and within acceptable physical limits, good concordance and excellent representativeness to the analyzed cases. It evidences that the differential formalism is an important tool to the sensitivity analysis and also it validates the application of the methodology in heat transmission problems on extended surfaces. The method proves to be necessary and efficient while elaborating thermal engineering projects. (author)

  8. Diffusive heat transport across magnetic islands and stochastic layers in tokamaks

    International Nuclear Information System (INIS)

    Hoelzl, Matthias

    2010-01-01

    Heat transport in tokamak plasmas with magnetic islands and ergodic field lines was simulated at realistic plasma parameters in realistic tokamak geometries. This requires the treatment of anisotropic heat diffusion, which is more efficient along magnetic field lines by up to ten orders of magnitude than perpendicular to them. Comparisons with analytical predictions and experimental measurements allow to determine the stability properties of neoclassical tearing modes as well as the experimental heat diffusion anisotropy.

  9. Major Variations in Sub-Tropical North Atlantic Heat Transport at Short Timescales: Causes and Consequences

    Science.gov (United States)

    Moat, Ben; Josey, Simon; Sinha, Bablu; Blaker, Adam; Smeed, David; McCarthy, Gerard; Johns, William; Hirschi, Joel; Frajka-Williams, Eleanor; Rayner, Darren; Duchez, Aurelie; Coward, Andrew

    2017-04-01

    Variability in the North Atlantic ocean heat transport at 26.5°N on short (5-day) timescales is identified and contrasted with different behaviour at monthly intervals using a combination of RAPID/MOCHA/WBTS measurements and the NEMO-LIM2 1/12° ocean circulation/sea ice model. Wind forcing plays the leading role in establishing the heat transport variability through the Ekman transport response of the ocean and the associated driving atmospheric conditions vary significantly with timescale. We find that at 5-day timescales the largest changes in the heat transport across 26.5N coincide with north-westerly airflows originating over the American land mass that drive strong southward anomalies in the Ekman flow. During these events the northward heat transport reduces by 0.5-0.7 PW (i.e. about 50% of the mean). In contrast, the Ekman transport response at longer monthly timescales is smaller in magnitude (0.2-0.3 PW) and consistent with expected variations in the leading mode of North Atlantic atmospheric variability, the North Atlantic Oscillation. The north-westerly airflow mechanism can have a prolonged influence beyond the central 5-day timescale and on occasion can reduce the accumulated winter ocean heat transport into the north Atlantic by 40%.

  10. Major variations in subtropical North Atlantic heat transport at short (5 day) timescales and their causes

    Science.gov (United States)

    Moat, B. I.; Josey, S. A.; Sinha, B.; Blaker, A. T.; Smeed, D. A.; McCarthy, G. D.; Johns, W. E.; Hirschi, J. J.-M.; Frajka-Williams, E.; Rayner, D.; Duchez, A.; Coward, A. C.

    2016-05-01

    Variability in the North Atlantic ocean heat transport at 26.5°N on short (5 day) timescales is identified and contrasted with different behaviour at monthly intervals using a combination of RAPID/MOCHA/WBTS measurements and the NEMO-LIM2 1/12° ocean circulation/sea ice model. Wind forcing plays the leading role in establishing the heat transport variability through the Ekman transport response of the ocean and the associated driving atmospheric conditions vary significantly with timescale. We find that at 5 day timescales the largest changes in the heat transport across 26.5°N coincide with north-westerly airflows originating over the American land mass that drive strong southward anomalies in the Ekman flow. During these events the northward heat transport reduces by 0.5-1.4 PW. In contrast, the Ekman transport response at longer monthly timescales is smaller in magnitude (up to 0.5 PW) and consistent with expected variations in the leading mode of North Atlantic atmospheric variability, the North Atlantic Oscillation. The north-westerly airflow mechanism can have a prolonged influence beyond the central 5 day timescale and on occasion can reduce the accumulated winter ocean heat transport into the North Atlantic by ˜40%.

  11. Nuclear reactor cavity floor passive heat removal system

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  12. The Sydvaerme project: District heating from the Barsebeck nuclear power plant

    International Nuclear Information System (INIS)

    Josefsson, L.

    1977-01-01

    The paper presents a summary report of a study on district heating from Barsebeck Nuclear Power Plant in Sweden, prepared cooperatively by the cities of Malmoe, Lund, Helsingborg, Landskrona and the electric power company Sydkraft. A future number 3 generating set at the Barsebeck nuclear power station could be designed for combined production of heat and electric power. The generating set could be completed after 1983, and could then supply about 65% of total district heating requirements. The first stage of the investigation includes a proposal for a technically feasible solution, sufficiently detailed to permit both technical and economic evaluation of the project. (author)

  13. Integrated system of nuclear reactor and heat exchanger

    International Nuclear Information System (INIS)

    McDonald, B.N.; Schluderberg, D.C.

    1977-01-01

    The invention concerns PWRs in which the heat exchanger is associated with a pressure vessel containing the core and from which it can be selectively detached. This structural configuration applies to electric power generating uses based on land or on board ships. An existing reactor of this kind is fitted with a heat exchanger in which the tubes are 'U' shaped. This particular design of heat exchangers requires that the ends of the curved tubes be solidly maintained in a tube plate of great thickness, hence difficult to handle and to fabricate and requiring unconventional fine control systems for the control rods and awkward coolant pump arrangements. These complications limit the thermal power of the system to level below 100 megawatts. On the contrary, the object of this invention is to provide a one-piece PWR reactor capable of reaching power levels of 1500 thermal megawatts at least. For this, a pressure vessel is provided in the cylindrical assembly with not only a transversal separation on a plane located between the reactor and the heat exchanger but also a cover selectively detachable which supports the fine control gear of the control rods. Removing the cover exposes a part of the heat exchanger for easy inspection and maintenance. Further, the heat exchanger can be removed totally from the pressure vessel containing the core by detaching the cylindrical part, which composes the heat exchanger section, from the part that holds the reactor core on a level with the transversal separation [fr

  14. Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

    International Nuclear Information System (INIS)

    1995-03-01

    This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

  15. VS2DRTI: Simulating Heat and Reactive Solute Transport in Variably Saturated Porous Media.

    Science.gov (United States)

    Healy, Richard W; Haile, Sosina S; Parkhurst, David L; Charlton, Scott R

    2018-01-29

    Variably saturated groundwater flow, heat transport, and solute transport are important processes in environmental phenomena, such as the natural evolution of water chemistry of aquifers and streams, the storage of radioactive waste in a geologic repository, the contamination of water resources from acid-rock drainage, and the geologic sequestration of carbon dioxide. Up to now, our ability to simulate these processes simultaneously with fully coupled reactive transport models has been limited to complex and often difficult-to-use models. To address the need for a simple and easy-to-use model, the VS2DRTI software package has been developed for simulating water flow, heat transport, and reactive solute transport through variably saturated porous media. The underlying numerical model, VS2DRT, was created by coupling the flow and transport capabilities of the VS2DT and VS2DH models with the equilibrium and kinetic reaction capabilities of PhreeqcRM. Flow capabilities include two-dimensional, constant-density, variably saturated flow; transport capabilities include both heat and multicomponent solute transport; and the reaction capabilities are a complete implementation of geochemical reactions of PHREEQC. The graphical user interface includes a preprocessor for building simulations and a postprocessor for visual display of simulation results. To demonstrate the simulation of multiple processes, the model is applied to a hypothetical example of injection of heated waste water to an aquifer with temperature-dependent cation exchange. VS2DRTI is freely available public domain software. © 2018, National Ground Water Association.

  16. Dynamics of water transport and storage in conifers studied with deuterium and heat tracing techniques.

    Science.gov (United States)

    F.C. Meinzer; J.R. Brooks; J.-C. Domec; B.L. Gartner; J.M. Warren; D.R. Woodruff; K. Bible; D.C. Shaw

    2006-01-01

    The volume and complexity of their vascular systems make the dynamics of tong-distance water transport in large trees difficult to study. We used heat and deuterated water (D20) as tracers to characterize whole-tree water transport and storage properties in individual trees belonging to the coniferous species Pseudotsuga menziesii...

  17. Physical protection of export/import and transportation of nuclear material in the Slovak Republic

    International Nuclear Information System (INIS)

    Vaclav, J

    2002-01-01

    Full text: The paper contains short overview about average amount of nuclear materials transported on the territory of the Slovak Republic in a year, and the physical protection of these nuclear materials. There are several types of transportation and export/import of nuclear materials in the SR: fresh fuel import; import of other unirradiated nuclear materials (e.g. depleted uranium, natural uranium); export of unirradiated nuclear materials (e.g. natural uranium); internal transportation of fresh fuel; internal transportation of other unirradiated nuclear materials; internal transportation of spent fuel. The main objective of the nuclear regulatory authority SR is to supervise observation of the national legislation as follows: the act no. 130 / 1998 on peaceful use of nuclear energy; UJD SR's regulation no. 186/1999 which details the physical protection of the nuclear facilities, nuclear materials, and radioactive waste (following requirements of INFCIRC 225 / Rev. 4); UJD SR's regulation no. 284 / 1999 which details conditions of nuclear material and radioactive wastes transportation. (author)

  18. Spent nuclear fuel transportation: public issues and answers

    International Nuclear Information System (INIS)

    Hoffman, W.D.

    1986-01-01

    The court-ordered shipping of 750 spent nuclear fuel assemblies from West Valley, New York back to their utility owners has generated considerable public and media interest. This paper discusses the specific concerns of the general public over the West Valley shipments, the issues raised by opposition groups, the interest of public officials and emergency preparedness teams as well as the media coverage generated. An analysis is performed on the effectiveness of the West Valley and utility public information programs utilized in addressing these issues, concerns and interests. Emphasis is placed on communications which work to facilitate the shipments and generate fuel transport acceptance. Information programs are discussed which increase preparedness for nuclear shipments by emergency response teams and build public confidence in their safety. The paper also examines communications which could have further enhanced the shipping campaign to date. Finally, plans are discussed for media preparation with interview training and press conferences. Emphasis is placed on materials provided for the media which have served to generate more favorable print and air time

  19. Nuclear process heat at high temperature: Application, realization and development programme

    International Nuclear Information System (INIS)

    Sammeck, K.H.; Fischer, R.

    1976-01-01

    Studies in the Federal Republic of Germany (FRG), the USA and the United Kingdom have shown that high-temperature helium energy from an HTR can advantageously be utilized for coal gasification and other fossil fuel conversion processes, and that a substantial demand for substitute natural gas (SNG) can be expected in the future. These results are based on plant design studies, economic assessments and basic development efforts in the field of coal gasification with nuclear heat, which in the FRG were carried out by Arbeitsgemeinschaft Nukleare Prozesswaerme (ANP)-members, HRB and KFA Juelich. Nuclear process plants are based on different gasification processes, resulting in different concepts of the nuclear heat system. In the case of hydro-gasification it is expected that steam reformers, arranged within the primary circuit of the reactor, will be heated directly by the primary helium. In the case of steam gasification, the high-temperature energy must be transferred to the gasification process via an intermediate circuit which is coupled to a gasifier outside the containment. In both cases the design of the nuclear reactor resembles an HTR for electricity generation. The main objectives of the development of nuclear process heat are to increase the helium outlet temperature of the reactor up to 950 0 C, to develop metallic alloys for high-temperature components such as heat exchangers, to design and construct a hot-gas duct, a steam reformer and a helium-helium heat exchanger and to develop the gasification processes. The nuclear safety regulations and the interface problems between the reactor, the process plant and the electricity generating plant have to be considered thoroughly. The Arbeitsgemeinschaft Nukleare Prozesswaerme and HRB started a development programme, in close collaboration with KFA Juelich, which will lead to the construction of a prototype plant for coal gasification with nuclear heat within 5 to 5 1/2 years. A survey of the main objectives

  20. Heat pipe effects in nuclear waste isolation: a review

    International Nuclear Information System (INIS)

    Doughty, C.; Pruess, K.

    1985-12-01

    The existence of fractures favors heat pipe development in a geologic repository as does a partially saturated medium. A number of geologic media are being considered as potential repository sites. Tuff is partially saturated and fractured, basalt and granite are saturated and fractured, salt is unfractured and saturated. Thus the most likely conditions for heat pipe formation occur in tuff while the least likely occur in salt. The relative permeability and capillary pressure dependences on saturation are of critical importance for predicting thermohydraulic behavior around a repository. Mineral redistribution in heat pipe systems near high-level waste packages emplaced in partially saturated formations may significantly affect fluid flow and heat transfer processes, and the chemical environment of the packages. We believe that a combined laboratory, field, and theoretical effort will be needed to identify the relevant physical and chemical processes, and the specific parameters applicable to a particular site. 25 refs., 1 fig

  1. Role of ocean heat transport in climates of tidally locked exoplanets around M dwarf stars

    Science.gov (United States)

    Hu, Yongyun; Yang, Jun

    2014-01-01

    The distinctive feature of tidally locked exoplanets is the very uneven heating by stellar radiation between the dayside and nightside. Previous work has focused on the role of atmospheric heat transport in preventing atmospheric collapse on the nightside for terrestrial exoplanets in the habitable zone around M dwarfs. In the present paper, we carry out simulations with a fully coupled atmosphere–ocean general circulation model to investigate the role of ocean heat transport in climate states of tidally locked habitable exoplanets around M dwarfs. Our simulation results demonstrate that ocean heat transport substantially extends the area of open water along the equator, showing a lobster-like spatial pattern of open water, instead of an “eyeball.” For sufficiently high-level greenhouse gases or strong stellar radiation, ocean heat transport can even lead to complete deglaciation of the nightside. Our simulations also suggest that ocean heat transport likely narrows the width of M dwarfs’ habitable zone. This study provides a demonstration of the importance of exooceanography in determining climate states and habitability of exoplanets. PMID:24379386

  2. Role of ocean heat transport in climates of tidally locked exoplanets around M dwarf stars.

    Science.gov (United States)

    Hu, Yongyun; Yang, Jun

    2014-01-14

    The distinctive feature of tidally locked exoplanets is the very uneven heating by stellar radiation between the dayside and nightside. Previous work has focused on the role of atmospheric heat transport in preventing atmospheric collapse on the nightside for terrestrial exoplanets in the habitable zone around M dwarfs. In the present paper, we carry out simulations with a fully coupled atmosphere-ocean general circulation model to investigate the role of ocean heat transport in climate states of tidally locked habitable exoplanets around M dwarfs. Our simulation results demonstrate that ocean heat transport substantially extends the area of open water along the equator, showing a lobster-like spatial pattern of open water, instead of an "eyeball." For sufficiently high-level greenhouse gases or strong stellar radiation, ocean heat transport can even lead to complete deglaciation of the nightside. Our simulations also suggest that ocean heat transport likely narrows the width of M dwarfs' habitable zone. This study provides a demonstration of the importance of exooceanography in determining climate states and habitability of exoplanets.

  3. Pressurizer level measurement inside PWR nuclear plant using resistance type heat sensors

    International Nuclear Information System (INIS)

    El Moussaoui, Ahmed.

    1982-06-01

    The accident that occured in 1979 to the PWR type nuclear reactor, Three-Mile Island 2, has drawn attention to the maladjustement of the differentiel pressure level detector installed in nuclear plants on the market. A system is presented here for measuring the level in pressurizers based on measurements of the heat resistance of the boundary layer existing between the heated sensor and the fluid mass in the vessel. The sensor consists of a 3 cm diameter cylindrical insulator support around which a 0.1 mm diameter platinum filament is wound. This filament simultaneously fulfills heating and transducer functions. To verify the feasibility of the resistant type heat sensor a test system, which provides water and steam under pressure was realised. Static and dynamic tests have shown that the principle of the resistant heat sensor is viable and can be used to obtain level informations [fr

  4. Thickness Optimisation of Textiles Subjected to Heat and Mass Transport during Ironing

    Directory of Open Access Journals (Sweden)

    Korycki Ryszard

    2016-09-01

    Full Text Available Let us next analyse the coupled problem during ironing of textiles, that is, the heat is transported with mass whereas the mass transport with heat is negligible. It is necessary to define both physical and mathematical models. Introducing two-phase system of mass sorption by fibres, the transport equations are introduced and accompanied by the set of boundary and initial conditions. Optimisation of material thickness during ironing is gradient oriented. The first-order sensitivity of an arbitrary objective functional is analysed and included in optimisation procedure. Numerical example is the thickness optimisation of different textile materials in ironing device.

  5. Constraining nuclear data via cosmological observations: Neutrino energy transport and big bang nucleosynthesis

    Directory of Open Access Journals (Sweden)

    Paris Mark

    2017-01-01

    Full Text Available We introduce a new computational capability that moves toward a self-consistent calculation of neutrino transport and nuclear reactions for big bang nucleosynthesis (BBN. Such a self-consistent approach is needed to be able to extract detailed information about nuclear reactions and physics beyond the standard model from precision cosmological observations of primordial nuclides and the cosmic microwave background radiation. We calculate the evolution of the early universe through the epochs of weak decoupling, weak freeze-out and big bang nucleosynthesis (BBN by simultaneously coupling a full strong, electromagnetic, and weak nuclear reaction network with a multi-energy group Boltzmann neutrino energy transport scheme. The modular structure of our approach allows the dissection of the relative contributions of each process responsible for evolving the dynamics of the early universe. Such an approach allows a detailed account of the evolution of the active neutrino energy distribution functions alongside and self-consistently with the nuclear reactions and entropy/heat generation and 'ow between the neutrino and photon/electron/positron/baryon plasma components. Our calculations reveal nonlinear feedback in the time evolution of neutrino distribution functions and plasma thermodynamic conditions. We discuss the time development of neutrino spectral distortions and concomitant entropy production and extraction from the plasma. These e↑ects result in changes in the computed values of the BBN deuterium and helium-4 yields that are on the order of a half-percent relative to a baseline standard BBN calculation with no neutrino transport. This is an order of magnitude larger e↑ect than in previous estimates. For particular implementations of quantum corrections in plasma thermodynamics, our calculations show a 0.4% increase in deuterium and a 0.6% decrease in 4He over our baseline. The magnitude of these changes are on the order of uncertainties

  6. Transport phenomena in capillary-porous structures and heat pipes

    CERN Document Server

    Smirnov, Henry

    2009-01-01

    With emphasis on the processes involved, this text explores the experimental efforts in two-phase thermal control technology research and development. This work evaluates and compares different theoretical approaches, experimental results, and models, such as semi-empirical models for critical boiling heat fluxes.

  7. Evaluation of the transport and resuspension of a simulated nuclear waste slurry: Nuclear Waste Treatment Program

    Energy Technology Data Exchange (ETDEWEB)

    Carleson, T.E.; Drown, D.C.; Hart, R.E.; Peterson, M.E.

    1987-09-01

    The Department of Chemical Engineering at the University of Idaho conducted research on the transport and resuspension of a simulated high-level nuclear waste slurry. In the United States, the reference process for treating both defense and civilian HLLW is vitrification using the liquid-fed ceramic melter process. The non-Newtonian behavior of the slurry complicates the evaluation of the transport and resuspension characteristics of the slurry. The resuspension of a simulated (nonradioactive) melter feed slurry was evaluated using a slurry designated as WV-205. The simulated slurry was developed for the West Valley Demonstration Project and was used during a pilot-scale ceramic melter (PSCM) experiment conducted at PNL in July 1985 (PSCM-21). This study involved determining the transport characteristics of a fully suspended slurry and the resuspension characteristics of settled solids in a pilot-scale pipe loop. The goal was to predict the transport and resuspension of a full-scale system based on rheological data for a specific slurry. The rheological behavior of the slurry was evaluated using a concentric cylinder rotational viscometer, a capillary tube viscometer, and the pilot-scale pipe loop. The results obtained from the three approaches were compared. 40 refs., 74 figs., 15 tabs.

  8. Evaluation of the transport and resuspension of a simulated nuclear waste slurry: Nuclear Waste Treatment Program

    International Nuclear Information System (INIS)

    Carleson, T.E.; Drown, D.C.; Hart, R.E.; Peterson, M.E.

    1987-09-01

    The Department of Chemical Engineering at the University of Idaho conducted research on the transport and resuspension of a simulated high-level nuclear waste slurry. In the United States, the reference process for treating both defense and civilian HLLW is vitrification using the liquid-fed ceramic melter process. The non-Newtonian behavior of the slurry complicates the evaluation of the transport and resuspension characteristics of the slurry. The resuspension of a simulated (nonradioactive) melter feed slurry was evaluated using a slurry designated as WV-205. The simulated slurry was developed for the West Valley Demonstration Project and was used during a pilot-scale ceramic melter (PSCM) experiment conducted at PNL in July 1985 (PSCM-21). This study involved determining the transport characteristics of a fully suspended slurry and the resuspension characteristics of settled solids in a pilot-scale pipe loop. The goal was to predict the transport and resuspension of a full-scale system based on rheological data for a specific slurry. The rheological behavior of the slurry was evaluated using a concentric cylinder rotational viscometer, a capillary tube viscometer, and the pilot-scale pipe loop. The results obtained from the three approaches were compared. 40 refs., 74 figs., 15 tabs

  9. Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

    Science.gov (United States)

    Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

    2014-01-07

    A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

  10. Fluctuation theory for transport properties in multicomponent mixtures: thermodiffusion and heat conductivity

    DEFF Research Database (Denmark)

    Shapiro, Alexander

    2004-01-01

    The theory of transport properties in multicomponent gas and liquid mixtures, which was previously developed for diffusion coefficients, is extended onto thermodiffusion coefficients and heat conductivities. The derivation of the expressions for transport properties is based on the general statis...... of the heat conductivity coefficient for ideal gas. (C) 2003 Elsevier B.V. All rights reserved.......The theory of transport properties in multicomponent gas and liquid mixtures, which was previously developed for diffusion coefficients, is extended onto thermodiffusion coefficients and heat conductivities. The derivation of the expressions for transport properties is based on the general...... statistical theory of fluctuations around an equilibrium state. The Onsager matrix of phenomenological coefficients is expressed in terms of the penetration lengths, including the newly introduced penetration length for the energy transfer. As an example, this penetration length is found from the known value...

  11. Study of Heat Flux Threshold and Perturbation Effect on Transport Barrier Formation Based on Bifurcation Model

    International Nuclear Information System (INIS)

    Chatthong, B.; Onjun, T.; Imbeaux, F.; Sarazin, Y.; Strugarek, A.; Picha, R.; Poolyarat, N.

    2011-06-01

    Full text: Formation of transport barrier in fusion plasma is studied using a simple one-field bistable S-curve bifurcation model. This model is characterized by an S-line with two stable branches corresponding to the low (L) and high (H) confinement modes, connected by an unstable branch. Assumptions used in this model are such that the reduction in anomalous transport is caused by v E velocity shear effect and also this velocity shear is proportional to pressure gradient. In this study, analytical and numerical approaches are used to obtain necessary conditions for transport barrier formation, i.e. the ratio of anomalous over neoclassical coefficients and heat flux thresholds which must be exceeded. Several profiles of heat sources are considered in this work including constant, Gaussian, and hyperbolic tangent forms. Moreover, the effect of perturbation in heat flux is investigated with respect to transport barrier formation

  12. Graphene transport properties upon exposure to PMMA processing and heat treatments

    DEFF Research Database (Denmark)

    Gammelgaard, Lene; Caridad, Jose; Cagliani, Alberto

    2014-01-01

    The evolution of graphene's electrical transport properties due to processing with the polymer polymethyl methacrylate (PMMA) and heat are examined in this study. The use of stencil (shadow mask) lithography enables fabrication of graphene devices without the usage of polymers, chemicals or heat......, allowing us to measure the evolution of the electrical transport properties during individual processing steps from the initial as-exfoliated to the PMMA-processed graphene. Heating generally promotes the conformation of graphene to SiO2 and is found to play a major role for the electrical properties...... that flakes conforming poorly to the substrate will have a higher carrier mobility which will however be reduced as heat treatment enhance the conformation. We finally show the electrical properties of graphene to be reversible upon heat treatments in air up to 200°C....

  13. Thermal-Hydraulic Analyses of Heat Transfer Fluid Requirements and Characteristics for Coupling A Hydrogen Production Plant to a High-Temperature Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis; C. H. Oh; R. B. Barner; D. F. Wilson

    2005-06-01

    The Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the hightemperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant, may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. Seven possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermalhydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermalhydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The

  14. Characteristics of nonlocally-coupled transition of the heat transport in LHD

    International Nuclear Information System (INIS)

    Tamura, N.; Ida, K.; Tanaka, K.; Tokuzawa, T.; Itoh, K.; Shimozuma, T.; Kubo, S.; Tsuchiya, H.; Nagayama, Y.; Kawahata, K.; Sudo, S.; Yamada, H.; Inagaki, S.

    2010-01-01

    A comparison of characteristics between a nonlocal transport phenomenon and an electron internal transport barrier (ITB) in the Large Helical Device is performed with a transient transport analysis and from the viewpoint of a dynamic behavior of transport state. The electron ITB is characterized by a jump of electron temperature gradient. In contrast, the transient transport analysis indicates the nonlocal transport phenomenon is characterized by a jump of electron heat flux. And seen from the viewpoint of the dynamic behavior of transport state, the physical mechanism of the appearance of the nonlocal transport phenomenon is found to be qualitatively different from that of the formation of the electron ITB. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  15. 10 CFR 70.20a - General license to possess special nuclear material for transport.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license to possess special nuclear material for transport. 70.20a Section 70.20a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DOMESTIC LICENSING OF... transport. (a) A general license is issued to any person to possess formula quantities of strategic special...

  16. Regulatory Framework for the Safe and Secure Transport of Nuclear Material in Japan

    International Nuclear Information System (INIS)

    Konnai, A.; Shibasaki, N.; Ikoma, Y.; Kato, M.; Yamauchi, T.; Iwasa, T.

    2016-01-01

    Regulations for nuclear material transport in Japan are based on international regulations. Safety and security regulations, however, have sometime different aspects which have caused a conflict of operations. This paper aims to introduce framework of safety and security regulations for nuclear material transport in Japan, and shows some issues in cooperation of these regulations. (author)

  17. 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)

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

    International Nuclear Information System (INIS)

    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 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)

  19. Integrated heat transport simulation of high ion temperature plasma of LHD

    International Nuclear Information System (INIS)

    Murakami, S.; Yamaguchi, H.; Sakai, A.

    2014-10-01

    A first dynamical simulation of high ion temperature plasma with carbon pellet injection of LHD is performed by the integrated simulation GNET-TD + TASK3D. NBI heating deposition of time evolving plasma is evaluated by the 5D drift kinetic equation solver, GNET-TD and the heat transport of multi-ion species plasma (e, H, He, C) is studied by the integrated transport simulation code, TASK3D. Achievement of high ion temperature plasma is attributed to the 1) increase of heating power per ion due to the temporal increase of effective charge, 2) reduction of effective neoclassical transport with impurities, 3) reduction of turbulence transport. The reduction of turbulence transport is most significant contribution to achieve the high ion temperature and the reduction of the turbulent transport from the L-mode plasma (normal hydrogen plasma) is evaluated to be a factor about five by using integrated heat transport simulation code. Applying the Z effective dependent turbulent reduction model we obtain a similar time behavior of ion temperature after the C pellet injection with the experimental results. (author)

  20. Modification of the finite element heat and mass transfer code (FEHMN) to model multicomponent reactive transport

    International Nuclear Information System (INIS)

    Viswanathan, H.S.

    1995-01-01

    The finite element code FEHMN is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developed hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent K d model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also provide that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies

  1. Nuclear power and heating plants in the electric power system. Part I

    International Nuclear Information System (INIS)

    Kalincik, L.

    1975-01-01

    Procedures used and results obtained in the following works are described: Incorporation of the nuclear power plants in the power system in the long term perspective; physical limitations on the WWER 440 reactor power changes during fuel campaigns; evaluation of the consumption and start-up characteristics of WWER type nuclear power plants (2x440 MWe); evaluation of refuelling campaigns distribution of nuclear power plant units with regard to comprehensive control properties of nuclear power plants; the possibilities are investigated of the utilization of the WWER type reactor for heat supply in Czechoslovakia. (author)

  2. Nuclear ferritin in corneal epithelial cells: tissue-specific nuclear transport and protection from UV-damage.

    Science.gov (United States)

    Linsenmayer, Thomas F; Cai, Cindy X; Millholland, John M; Beazley, Kelly E; Fitch, John M

    2005-03-01

    We have identified the heavy chain of ferritin as a developmentally regulated nuclear protein of embryonic chicken corneal epithelial cells. The nuclear ferritin is assembled into a supramolecular form that is indistinguishable from the cytoplasmic form of ferritin found in other cell types. Thus it most likely has iron-sequestering capabilities. Free iron, via the Fenton reaction, is known to exacerbate UV-induced and other oxidative damage to cellular components, including DNA. Since corneal epithelial cells are constantly exposed to UV light, we hypothesized that the nuclear ferritin might protect the DNA of these cells from free radical damage. To test this possibility, primary cultures of cells from corneal epithelium and other tissues were UV irradiated, and damage to DNA was detected by an in situ 3'-end labeling assay. Consistent with the hypothesis, corneal epithelial cells with nuclear ferritin had significantly less DNA breakage than the other cells types examined. However, when the expression of nuclear ferritin was inhibited the cells now became much more susceptible to UV-induced DNA damage. Since ferritin is normally cytoplasmic, corneal epithelial cells must have a mechanism that effects its nuclear localization. We have determined that this involves a nuclear transport molecule which binds to ferritin and carries it into the nucleus. This transporter, which we have termed ferritoid for its similarity to ferritin, has at least two domains. One domain is ferritin-like and is responsible for binding the ferritin; the other domain contains a nuclear localization signal that is responsible for effecting the nuclear transport. Therefore, it seems that corneal epithelial cells have evolved a novel, nuclear ferritin-based mechanism for protecting their DNA against UV damage. In addition, since ferritoid is structurally similar to ferritin, it may represent an example of a nuclear transporter that evolved from the molecule it transports (i.e., ferritin).

  3. Evaluation of the Safety Issue Concerning the Potential for Loss of Decay Heat Removal Function due to Crude Oil Spill in the Ultimate Heat Sink of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Roh, Kyung Wan; Yune, Young Gill; Kang, Dong Gu; Kim, Hho Jhung [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    A barge crashed into a moored oil tanker at about 7:15 a.m., Dec. 12, 2007, dumping around 10,500 tons of crude oil into the sea in Korea. The incident took place about 15 kilometers northwest of Manripo beach in South Chungcheong where is Korea's west coast in the Yellow Sea. In a few days, the oil slicks spread to the northern and southern tips of the Taean Peninsula by strong winds and tides. As time went the spilled oil floating on the surface of sea water was volatilized to become tar-balls and lumps and drifted far away in the southern direction. 13 days after the incident, some of oil slicks and tar lumps were observed to flow in the service water intake at the Younggwang nuclear power plants (NPPs) operating 6 reactors, which are over 150 km away from the incident spot in the southeastern direction. According to the report by the Younggwang NPPs, a total weight 83 kg of tar lumps was removed for about 3 days. Oil spills in the sea can happen in any country or anytime due to human errors or mistakes, wars, terrors, intentional dumping of waste oils, and natural disasters like typhoon and tsunami. In fact, there have been 7 major oil spills over 10,000 tons that have occurred around the world since 1983. As such serious oil spill incidents may happen near the operating power plants using the sea water as ultimate heat sink. To ensure the safe operation of nuclear reactors it is required to evaluate the potential for loss of decay heat removal function of nuclear reactors due to the spilled oils flowing in the service water intake, from which the service water is pumped. Thus, Korea Institute of Nuclear Safety identified this problem as one of the important safety. When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where all service water pumps locate, by sea current and wind drift (induced) current. The essential service water pumps take the

  4. Nonlocal heat transport and improved target design for x-ray heating studies at x-ray free electron lasers

    Science.gov (United States)

    Hoidn, Oliver; Seidler, Gerald T.

    2018-01-01

    The extremely high-power densities and short durations of single pulses of x-ray free electron lasers (XFELs) have opened new opportunities in atomic physics, where complex excitation-relaxation chains allow for high ionization states in atomic and molecular systems, and in dense plasma physics, where XFEL heating of solid-density targets can create unique dense states of matter having temperatures on the order of the Fermi energy. We focus here on the latter phenomena, with special emphasis on the problem of optimum target design to achieve high x-ray heating into the warm dense matter (WDM) state. We report fully three-dimensional simulations of the incident x-ray pulse and the resulting multielectron relaxation cascade to model the spatial energy density deposition in multicomponent targets, with particular focus on the effects of nonlocal heat transport due to the motion of high energy photoelectrons and Auger electrons. We find that nanoscale high-Z /low-Z multicomponent targets can give much improved energy density deposition in lower-Z materials, with enhancements reaching a factor of 100. This has three important benefits. First, it greatly enlarges the thermodynamic parameter space in XFEL x-ray heating studies of lower-Z materials. Second, it allows the use of higher probe photon energies, enabling higher-information content x-ray diffraction (XRD) measurements such as in two-color XFEL operations. Third, while this is merely one step toward optimization of x-ray heating target design, the demonstration of the importance of nonlocal heat transport establishes important common ground between XFEL-based x-ray heating studies and more traditional laser plasma methods.

  5. Metallic materials for heat exchanger components and highly stressed internal of HTR reactors for nuclear process heat generation

    International Nuclear Information System (INIS)

    1982-01-01

    The programme was aimed at the development and improvement of materials for the high-temperature heat exchanger components of a process steam HTR. The materials must have high resistance to corrosion, i.e. carburisation and internal oxidation, and high long-term toughness over a wide range of temperatures. They must also meet the requirements set in the nuclear licensing procedure, i.e. resistance to cyclic stress and irradiation, non-destructive testing, etc. Initially, it was only intended to improve and qualify commercial alloys. Later on an alloy development programme was initiated in which new, non-commercial alloys were produced and modified for use in a nuclear process heat facility. Separate abstracts were prepared for 19 pays of this volume. (orig./IHOE) [de

  6. On-site transportation and handling of uranium-233 special nuclear material: Preliminary hazards and accident analysis. Final

    International Nuclear Information System (INIS)

    Solack, T.; West, D.; Ullman, D.; Coppock, G.; Cox, C.

    1995-01-01

    U-233 Special Nuclear Material (SNM) currently stored at the T-Building Storage Areas A and B must be transported to the SW/R Tritium Complex for repackaging. This SNM is in the form of oxide powder contained in glass jars which in turn are contained in heat sealed double polyethylene bags. These doubled-bagged glass jars have been primarily stored in structural steel casks and birdcages for approximately 20 years. The three casks, eight birdcages, and one pail/pressure vessel will be loaded onto a transport truck and moved over an eight day period. The Preliminary Hazards and Accident Analysis for the on-site transportation and handling of Uranium-233 Special Nuclear Material, documented herein, was performed in accordance with the format and content guidance of DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, dated July 1994, specifically Chapter Three, Hazard and Accident Analysis. The Preliminary Hazards Analysis involved detailed walkdowns of all areas of the U-233 SNM movement route, including the T-Building Storage Area A and B, T-Building truck tunnel, and the roadway route. Extensive discussions were held with operations personnel from the Nuclear Material Control Group, Nuclear Materials Accountability Group, EG and G Mound Security and the Material Handling Systems Transportation Group. Existing documentation related to the on-site transportation of hazardous materials, T-Building and SW/R Tritium Complex SARs, and emergency preparedness/response documentation were also reviewed and analyzed to identify and develop the complete spectrum of energy source hazards

  7. The Influence of Rain Sensible Heat and Subsurface Energy Transport on the Energy Balance at the Land Surface

    NARCIS (Netherlands)

    Kollet, S.J.; Cvijanovic, I.; Schüttemeyer, D.; Maxwell, R.M.; Moene, A.F.; Bayer, P.

    2009-01-01

    In land surface models, which account for the energy balance at the land surface, subsurface heat transport is an important component that reciprocally influences ground, sensible, and latent heat fluxes and net radiation. In most models, subsurface heat transport parameterizations are commonly

  8. Heat supply analysis of steam reforming hydrogen production process in conventional and nuclear

    International Nuclear Information System (INIS)

    Siti Alimah; Djati Hoesen Salimy

    2015-01-01

    Tile analysis of heat energy supply in the production of hydrogen by natural gas steam reforming process has been done. The aim of the study is to compare the energy supply system of conventional and nuclear heat. Methodology used in this study is an assessment of literature and analysis based on the comparisons. The study shows that the heat sources of fossil fuels (natural gas) is able to provide optimum operating conditions of temperature and pressure of 850-900 °C and 2-3 MPa, as well as the heat transfer is dominated by radiation heat transfer, so that the heat flux that can be achieved on the catalyst tube relatively high (50-80 kW/m 2 ) and provide high thermal efficiency of about 85 %. While in the system with nuclear energy, due to the demands of safety, process operating at less than optimum conditions of temperature and pressure of 800-850 °C and 4.5 MPa, as well as the heat transfer is dominated by convection heat transfer, so that the heat flux that can be achieved catalyst tube is relatively low (1020 kW/m 2 ) and it provides a low thermal efficiency of about 50 %. Modifications of reformer and heat utilization can increase the heat flux up to 40 kW/m 2 so that the thermal efficiency can reach 78 %. Nevertheless, the application of nuclear energy to hydrogen production with steam reforming process is able to reduce the burning of fossil fuels which has implications for the potential decrease in the rate of CO2 emissions into the environment. (author)

  9. Nonlinear charge transport in bipolar semiconductors due to electron heating

    Energy Technology Data Exchange (ETDEWEB)

    Molina-Valdovinos, S., E-mail: sergiom@fisica.uaz.edu.mx [Universidad Autónoma de Zacatecas, Unidad Académica de Física, Calzada Solidaridad esq. Paseo, La Bufa s/n, CP 98060, Zacatecas, Zac, México (Mexico); Gurevich, Yu.G. [Centro de Investigación y de Estudios Avanzados del IPN, Departamento de Física, Av. IPN 2508, México D.F., CP 07360, México (Mexico)

    2016-05-27

    It is known that when strong electric field is applied to a semiconductor sample, the current voltage characteristic deviates from the linear response. In this letter, we propose a new point of view of nonlinearity in semiconductors which is associated with the electron temperature dependence on the recombination rate. The heating of the charge carriers breaks the balance between generation and recombination, giving rise to nonequilibrium charge carriers concentration and nonlinearity. - Highlights: • A new mechanism of nonlinearity of current-voltage characteristic (CVC) is proposed. • The hot electron temperature violates the equilibrium between electrons and holes. • This violation gives rise to nonequilibrium concentration of electrons and holes. • This leads to nonlinear CVC (along with the heating nonlinearity).

  10. Photothermal heating in metal-embedded microtools for material transport

    DEFF Research Database (Denmark)

    Villangca, Mark Jayson; Palima, Darwin; Banas, Andrew Rafael

    2016-01-01

    Material transport is an important mechanism in microfluidics and drug delivery. The methods and solutions found in literature involve passively diffusing structures, microneedles and chemically fueled structures. In this work, we make use of optically actuated microtools with embedded metal layer...

  11. Heat and momentum transport in sheared Rayleigh-Benard convection

    Science.gov (United States)

    Krishnamurti, Ruby; Zhu, Yi

    This is a report on laboratory measurements of heat and momentum flux in sheared Rayleigh-Benard convection. A cylindrical annulus of water uniformly heated below and cooled above was subjected to an imposed shear by driving the bottom boundary so that it rotated steadily around the vertical axis of symmetry. The top boundary was stationary and formed part of a torsion balance. It was suspended from the laboratory ceiling by a long torsion wire. The torque exerted upon the top boundary by the fluid was measured by the twist of the torsion wire. From this angle of twist, the vertical flux of horizontal momentum was determined. The heat flux as well as the momentum flux were measured at each fixed value of Rayleigh number Ra and Reynolds number Re. The measurements were made in the range 6 × 10 6 ≤ Ra ≤ 6 × 10 7 and 38 ≤ Re ≤ 352. One of the most striling results is that at Ra near 3 × 10 7 to 5 × 10 7 depending upon Re, the dimensionless momentum flux Mo ceases its increasing trend and begins to decrease with increasing Ra. This may be described by saying that the effective viscosity decreases with further increasing of Ra. However, Mo always remained greater than unity in the range investigated.

  12. Legal and Institutional Issues of Transportable Nuclear Power Plants: A Preliminary Study

    International Nuclear Information System (INIS)

    2013-01-01

    jointly the international and national actions required for ensuring the sustainability of nuclear energy through innovations in technology and/or institutional arrangements. A transportable nuclear power plant (TNPP) is a factory manufactured, transportable and relocatable nuclear power plant which, when fuelled, is capable of producing final energy products such as electricity and heat. Introducing a TNPP may require fewer financial and human resources from the host State. However, the deployment of such reactors will face new legal issues in the international context which need to be resolved to enable the deployment of such reactors in countries other than the country of origin. The objective of this report is to study the legal and institutional issues for the deployment of TNPPs, to reveal challenges that might be faced in their deployment, and to outline pathways for resolution of the identified issues and challenges in the short and long terms. It is addressed to senior legal, regulatory and technical officers in Member States planning to embark on a nuclear power programme or to expand an existing one by considering the introduction of a TNPP

  13. Japan's regulatory and safety issues regarding nuclear materials transport

    Energy Technology Data Exchange (ETDEWEB)

    Saito, T. [Nuclear and Industrial Safety Agency, Ministry of Economy, Trade and Industry, Government of Japan, Tokyo (Japan); Yamanaka, T. [Japan Nuclear Energy Safety Organization, Government of Japan, Tokyo (Japan)

    2004-07-01

    This paper focuses on the regulatory and safety issues on nuclear materials transport which the Government of Japan (GOJ) faces and needs to well handle. Background information about the status of nuclear power plants (NPP) and nuclear fuel cycle (NFC) facilities in Japan will promote a better understanding of what this paper addresses.

  14. Nuclear source of district heating in the north-east region of Russia

    International Nuclear Information System (INIS)

    Dolgov, V.V.

    1998-01-01

    The operation of the Bilibin Nuclear Co-generation Plant (BNCP) as a local district heating source is reviewed in this paper. Specific features of the BNCP power unit are given with special emphases on the components of the technological scheme, which are involved in the heat production and supply to the consumers. The scheme of steam extraction from the turbine, the flow diagram of steam in the turbine, as well as the three circuit heat removal system are described. The numerical characteristics of the nuclear heat supply system in various operating modes are presented. The real information characterizing current radiological conditions in the vicinity of the heat generation and distribution equipment is also presented in the paper. The BNCP technical and economical characteristics are compared with those of conventional energy sources. Both advantages and some problems revealed during the twenty-year experience of the BNCP nuclear heat utilization are generally assessed. Safety and reliability characteristics of the reactor and the heat supply system are also described. (author)

  15. Shallow heat injection and storage experiment monitored with electrical resistivity tomography and simulated with heat transport model

    OpenAIRE

    Hermans, Thomas; Daoudi, Moubarak; Vandenbohede, Alexander; Lebbe, Luc; Nguyen, Frédéric

    2011-01-01

    Groundwater resources are increasingly used around the world as geothermal systems. Understanding physical processes and quantification of parameters determining heat transport in porous media is therefore important. To monitor the geothermal behavior of groundwater systems and to estimate the governing parameters, we rely mainly on borehole observations of the temperature field at a few locations (temperature logs or thermal response test). In analogy to research in hydrogeophysics, geophysi...

  16. Analysis for Heat Transfer in a High Current-Passing Carbon Nanosphere Using Nontraditional Thermal Transport Model.

    Science.gov (United States)

    Hol C Y; Chen, B C; Tsai, Y H; Ma, C; Wen, M Y

    2015-11-01

    This paper investigates the thermal transport in hollow microscale and nanoscale spheres subject to electrical heat source using nontraditional thermal transport model. Working as supercapacitor electrodes, carbon hollow micrometer- and nanometer-sized spheres needs excellent heat transfer characteristics to maintain high specific capacitance, long cycle life, and high power density. In the nanoscale regime, the prediction of heat transfer from the traditional heat conduction equation based on Fourier's law deviates from the measured data. Consequently, the electrical heat source-induced heat transfer characteristics in hollow micrometer- and nanometer-sized spheres are studied using nontraditional thermal transport model. The effects of parameters on heat transfer in the hollow micrometer- and nanometer-sized spheres are discussed in this study. The results reveal that the heat transferred into the spherical interior, temperature and heat flux in the hollow sphere decrease with the increasing Knudsen number when the radius of sphere is comparable to the mean free path of heat carriers.

  17. 3D modeling of groundwater heat transport in the shallow Westliches Leibnitzer Feld aquifer, Austria

    Science.gov (United States)

    Rock, Gerhard; Kupfersberger, Hans

    2018-02-01

    For the shallow Westliches Leibnitzer feld aquifer (45 km2) we applied the recently developed methodology by Kupfersberger et al. (2017a) to derive the thermal upper boundary for a 3D heat transport model from observed air temperatures. We distinguished between land uses of grass and agriculture, sealed surfaces, forest and water bodies. To represent the heat flux from heated buildings and the mixture between different land surfaces in urban areas we ran the 1D vertical heat conduction module SoilTemp which is coupled to the heat transport model (using FEFLOW) on a time step basis. Over a simulation period of 23 years the comparison between measured and observed groundwater temperatures yielded NSE values ranging from 0.41 to 0.92 including readings at different depths. The model results showed that the thermal input signals lead to distinctly different vertical groundwater temperature distributions. To overcome the influence of specific warm or cold years we introduced the computation of an annual averaged groundwater temperature profile. With respect to the use of groundwater cooling or heating facilities we evaluated the application of vertically averaged statistical groundwater temperature distributions compared to the use of temperature distributions at selected dates. We concluded that the heat transport model serves well as an aquifer scale management tool to optimize the use of the shallow subsurface for thermal purposes and to analyze the impacts of corresponding measures on groundwater temperatures.

  18. Effect of rotational speed modulation on heat transport in a fluid layer with temperature dependent viscosity and internal heat source

    Directory of Open Access Journals (Sweden)

    B.S. Bhadauria

    2014-12-01

    Full Text Available In this paper, a theoretical investigation has been carried out to study the combined effect of rotation speed modulation and internal heating on thermal instability in a temperature dependent viscous horizontal fluid layer. Rayleigh–Bénard momentum equation with Coriolis term has been considered to describe the convective flow. The system is rotating about it is own axis with non-uniform rotational speed. In particular, a time-periodic and sinusoidally varying rotational speed has been considered. A weak nonlinear stability analysis is performed to find the effect of modulation on heat transport. Nusselt number is obtained in terms of amplitude of convection and internal Rayleigh number, and depicted graphically for showing the effects of various parameters of the system. The effect of modulated rotation speed is found to have a stabilizing effect for different values of modulation frequency. Further, internal heating and thermo-rheological parameters are found to destabilize the system.

  19. Current status of sea transport of nuclear fuel materials and LLW in Japan

    International Nuclear Information System (INIS)

    Kitagawa, Hiroshi; Akiyama, Hideo

    2000-01-01

    Along with the basic policy of the nuclear fuel cycle of Japan, many fuel cycle facilities have been already constructed in Rokkasho-Mura, Aomori prefecture, such as the uranium enrichment plant, the low level waste disposal center and the receiving pool of the spent nuclear fuels for reprocessing. These facilities belong to the Japan Nuclear Fuel Limited. (JNFL). Domestic sea transport of the spent nuclear fuels (SF) has been carried out since 1977 to the Tokai Reprocessing Plant, and the first sea transport of the SF to the fuel cycle facility in Rokkasho-Mura was done in Oct, 1998 using a new exclusive ship 'Rokuei-Maru'. Sea transport of the low level radioactive wastes (LLW) has been carried out since 1992 to the Rokkasho LLW Disposal Center, and about 130,000 LLW drams were transported from the nuclear power plant sites. These sea transport have demonstrated the safety of the transport of the nuclear fuel cycle materials. It is hoped that the safe sea transport of the nuclear fuel materials will contribute to the more progress of the nuclear fuel cycle activities of Japan. (author)

  20. Hygiene problems in building a nuclear power and heat plant near Bratislava

    International Nuclear Information System (INIS)

    Chorvat, D.; Mizov, J.; Hladky, E.; Kubik, I.; Carach, J.

    1976-01-01

    The results are presented of the calculation of the population exposure due to the release of radioactive products from a nuclear power and heating plant accident into the ambient atmosphere (primary coolant circuit rupture) providing the nuclear power and heating plant is sited approximately 500 m from the Slovnaft chemical works in Bratislava. Ground water contamination was analyzed assuming the infiltration of radioactive products from a surface deposit due to fallout and the direct infiltration of the products into the soil in the area of the plant. The results of the assessment of the design basis accident of a WWER-1000 nuclear power and heating plant show unequivocally that the emergency core cooling system, full-pressure containment and the correct function of the containment spray system are able to keep the accident consequences within acceptable limits thus meeting radiation hygiene requirements related to the siting of similar installations in the vicinity of large housing estates. (Oy)

  1. Assessment of irradiated nuclear spent fuel transport in the Commonwealth of independent states

    International Nuclear Information System (INIS)

    Raisonnier, D.; Lemogne, A.; Semin-Vadov, S.; Koshik, Y.

    1998-01-01

    Irradiated nuclear fuels (INFs) are transported regularly in and between the New Independent States (NIS). Under the TACIS Programme (Technical Assistance to the Commonwealth of Independent States), the European Commission has funded this project devoted to the assessment of transports of INF in Russia, Kazakhstan and Ukraine in collaboration with local State Organizations, competent Authorities, nuclear facilities and transport companies. This paper will present the development of this project during the two years of its implementation as well as its main results. (authors)

  2. Thermal ionization and plasma state of high temperature vapor of UO2, Cs, and Na: Effect on the heat and radiation transport properties of the vapor phase

    International Nuclear Information System (INIS)

    Karow, H.U.

    1979-01-01

    The paper deals with the question how far the thermophysical state and the convective and radiative heat transport properties of vaporized reactor core materials are affected by the thermal ionization existing in the actual vapor state. The materials under consideration here are: nuclear oxide fuel (UO 2 ), Na (as the LMFBR coolant material), and Cs (alkaline fission product, partly retained in the fuel of the core zone). (orig./RW) [de

  3. Application of thermohydraulic dispatcher in low temperature district heating systems for decreasing heat carrier transportation energy cost and increasing reliability of heat supply

    Science.gov (United States)

    Yavorovsky, Y. V.; Romanov, D. O.; Sennikov, V. V.; Sultanguzin, I. A.; Malenkov, A. S.; Zhigulina, E. V.; Lulaev, A. V.

    2017-11-01

    Low pressure district heating systems have low breakdown rate and allow decreasing heat carrier transportation energy cost by means of avoiding throttling of available water head. One of the basic elements of such systems is thermohydraulic dispatcher (THD) which separates primary circuit and secondary circuit (or circuits) that allows avoiding mutual hydraulic influence of circuits on each other and reducing water heads of network pumps. Analysis of perspective ways of using thermohydraulic dispatcher (THD) in low temperature district heating systems is made in this paper. Principal scheme and mathematical model of low pressure and temperature district heating system based on CHP generation with THD are considered. The main advantages of such systems are pointed out.

  4. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

    2009-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

  5. Fuel production from coal by the Mobil Oil process using nuclear high-temperature process heat

    International Nuclear Information System (INIS)

    Hoffmann, G.

    1982-01-01

    Two processes for the production of liquid hydrocarbons are presented: Direct conversion of coal into fuel (coal hydrogenation) and indirect conversion of coal into fuel (syngas production, methanol synthesis, Mobil Oil process). Both processes have several variants in which nuclear process heat may be used; in most cases, the nuclear heat is introduced in the gas production stage. The following gas production processes are compared: LURGI coal gasification process; steam reformer methanation, with and without coal hydrogasification and steam gasification of coal. (orig./EF) [de

  6. Some fabrication problems in nuclear power plants heat exchanges, its detectability and implications

    International Nuclear Information System (INIS)

    Condessa, N.C.; Oliveira, R.

    1988-01-01

    On the design and manufacturing follow-up of heat-exchangers of nuclear power plants some care are took into account in order to assure a high degree of confiability allowing the heat-exchanger in operation under severe and aggressive conditions be operating during the useful life of the nuclear power plant. However, despite the care, some problems can ocurr as the ones described on this job; that, if not detected in due time could bring umpleasant problems to the component or to the system in which it is working during operation. (author) [pt

  7. Heat-electricity convertion systems for a Brazilian space micro nuclear reactor

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine N.F.; Marcelino, Natalia B.; Placco, Guilherme M.; Nascimento, Jamil A.; Borges, Eduardo M.; Barrios Junior, Ary Garcia

    2013-01-01

    This contribution will discuss the evolution work in the development of thermal cycles to allow the development of heat-electricity conversion for the Brazilian space micro nuclear Reactor. Namely, innovative core and nuclear fuel elements, Brayton cycle, Stirling engine, heat pipes, passive multi-fluid turbine, among others. This work is basically to set up the experimental labs that will allow the specification and design of the space equipment. Also, some discussion of the cost so far, and possible other applications will be presented. (author)

  8. Economic evaluation of heat extraction from nuclear power plants - a criterion for deciding their building order

    International Nuclear Information System (INIS)

    Navratil, J.

    1987-01-01

    Heat extraction from nuclear power plants is an important element in the current concept of supplying the population and industries with heat. Economic evaluation of the extraction is one of the factors of the total economic assessment of potential sites for nuclear power plant construction which can contribute to decision making on the priorities of construction. The methodological approach to the assessment of economic contribution of heat extraction from 2x1000 MW nuclear power plant is exemplified using three such sites on the Czechoslovak territory, viz., Opatovice (eastern Bohemia), Blahutovice (northern Moravia), and Kecerovce (eastern Slovakia). The so-called annual converted cost was used as a suitable quantity completely reflecting all significant economic effects of heat extraction. It is shown that the fuel component of the power plant costs is the decisive factor for the amount of the annual converted cost in respect to heat supply and thus also the economic priority of the construction sites of nuclear power plants. (Z.M.). 3 tabs., 3 refs

  9. Dynamical transition of heat transport in a physical gel near the sol-gel transition

    Science.gov (United States)

    Kobayashi, Kazuya U.; Oikawa, Noriko; Kurita, Rei

    2015-12-01

    We experimentally study heat transport in a gelatin solution near a reversible sol-gel transition point where viscosity strongly depends on temperature. We visualize the temperature field and velocity field using thermochromic liquid crystals and polystyrene latex particles, respectively. During the initial stages of heating, we find that heat transport undergoes a dynamical transition from conductive to convective. Subsequently, during later stages, we observe that the transport dynamics are much more complex than conventional thermal convections. At the sample’s surface we observe the formation of stagnant domains, which lack fluid flow. Their formation is not due to the effects of local cooling. We determine that it is the dynamics of these stagnant domains that induce convective-conductive-convective transitions.

  10. 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.

  11. Thermal tolerances of fish from a reservoir receiving heated effluent from a nuclear reactor

    International Nuclear Information System (INIS)

    Holland, W.E.; Smith, M.H.; Gibbons, J.W.; Brown, D.H.

    1974-01-01

    The heat tolerances of bluegill (Lepomis macrochirus) subjected to heated effluent from a nuclear reactor was compared with those of bluegill living at normal temperatures. Three of the four study areas were located in the Par Pond reservoir system on the Savannah River Plant near Aiken, South Carolina. Results shown that at least one species of warm-water fish can adjust to elevated aquatic temperatures in a natural environment by becoming more tolerant. (U.S.)

  12. Study of a conceptual nuclear-energy center at Green River, Utah: site-specific transportation

    International Nuclear Information System (INIS)

    1981-10-01

    The objective of the following report is to assess the adequacy of the local and regional transportation network for handling traffic, logistics, and the transport of major power plant components to the Utah Nuclear Energy Center (UNEC) Horse Bench site. The discussion is divided into four parts: (1) system requirements; (2) description of the existing transportation network; (3) evaluation; (4) summary and conclusions

  13. The influence of meridional ice transport on Europa's ocean stratification and heat content

    Science.gov (United States)

    Zhu, P.; Manucharyan, G.; Thompson, A. F.; Goodman, J. C.; Vance, S.

    2017-12-01

    Jupiter's moon Europa likely hosts a saltwater ocean beneath its icy surface. Geothermal heating and rotating convection in the ocean may drive a global overturning circulation that redistributes heat vertically and meridionally, preferentially warming the ice shell at the equator. Here we assess thepreviously unconstrained influence of ocean-ice coupling on Europa's ocean stratification and heat transport. We demonstrate that a relatively fresh layer can form at the ice-ocean interface due to a meridional ice transport forced by the differential ice shell heating between the equator and the poles. We provide analytical and numerical solutions for the layer's characteristics, highlighting their sensitivity to critical ocean parameters. For a weakly turbulent and highly saline ocean, a strong buoyancy gradient at the base of the freshwater layer can suppress vertical tracer exchange with the deeper ocean. As a result, the freshwater layer permits relatively warm deep ocean temperatures.

  14. EFFECTIVENESS ANALYSIS OF CAMPUS HEAT SUPPLY SYSTEM OF DNIPROPETROVSK NATIONAL UNIVERSITY OF RAILWAY TRANSPORT

    Directory of Open Access Journals (Sweden)

    O. M. Pshinko

    2014-03-01

    Full Text Available Purpose. Heat consumption for heating and hot water supply of housing and industrial facilities is an essential part of heat energy consumption. Prerequisite for development of energy saving measures in existing heating systems is their preliminary examination. The investigation results of campus heating system of Dnipropetrovsk National University of Railway Transport named after Academician V. Lazaryan are presented in the article. On the basis of the analysis it is proposed to take the energy saving measures and assess their effectiveness. Methodology. Analysis of the consumption structure of thermal energy for heating domestic and hot water supply was fulfilled. The real costs of heat supply during the calendar year and the normative costs were compared. Findings. The recording expenditures data of thermal energy for heating supply of residential buildings and dormitories in 2012 were analyzed. The comparison of actual performance with specific regulations was performed. This comparison revealed problems, whose solution will help the efficient use of thermal energy. Originality. For the first time the impact of climate conditions, features of schemes and designs of heating systems on the effective use of thermal energy were analyzed. It was studied the contribution of each component. Practical value. Based on the analysis of thermal energy consumption it was developed a list of possible energy saving measures that can be implemented in the system of heat and power facilities. It was evaluated the fuel and energy resources saving.

  15. E-learning modules for nuclear reactor heat transfer

    Science.gov (United States)

    Jayaram, Praveen Bharadwaj

    E learning in engineering education is becoming popular at several universities as it allows instructors to create content that the students may view and interact with at his/her own convenience. Web-based simulation and what-if analysis are examples of such educational content and has proved to be extremely beneficial for engineering students. Such pedagogical content promote active learning and encourage students to experiment and be more creative. The main objective of this project is to develop web based learning modules, in the form of analytical simulations, for the Reactor Thermal Hydraulics course offered by the College of Engineering at UT Arlington. These modules seek to comprehensively transform the traditional education structure. The simulations are built to supplement the class lectures and are divided into categories like Fundamentals, Heat generation, Heat transfer and Heat removal categories. Each category contains modules which are sub-divided chapter wise and further into section wise. Some of the important sections from the text book are taken and calculations for a particular functionality are implemented. Since it is an interactive tool, it allows user to input certain values, which are then processed with the traditional equations, and output results either in the form of a number or graphs.

  16. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  17. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    International Nuclear Information System (INIS)

    Maraman, W.J.

    1979-12-01

    This formal monthly report covers the studies related to the use of 238 PuO 2 in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are general-purpose heat source development and space nuclear safety and fuels. Most of the studies discussed hear are of a continuing nature. Results and conclusions described may change as the work continues

  18. Study of two control rods of a district heating nuclear plant

    International Nuclear Information System (INIS)

    Martinez, J.M.

    1979-01-01

    This paper broaches the study of the control rods to ensure a convenient working during load following of the nuclear reactor THERMOS. The mathematical model is descriptive of the whole of the nuclear plant (point model for the core and the heat balances). Two power control are studied. The first, like PWR, is a program for the mean temperature of primary water. The second takes into account the structure of the plant and is described by a schedule of powers [fr

  19. Nuclear boiling heat transfer and critical heat flux in titanium dioxide-water nanofluids

    International Nuclear Information System (INIS)

    Okawa, Tomio; Takamura, Masahiro; Kamiya, Takahito

    2011-01-01

    Nucleate boiling heat transfer was experimentally studied for saturated pool boiling of water-based nanofluids. Since significant nanoparticle deposition on the heated surface was observed after the nucleate boiling in nanofluids, measurement of CHF was also carried out using the nanoparticle deposited heated surface; pure water was used in the CHF measurement. In the present work, the heated surface was a 20 mm diameter cupper surface, and titanium-dioxide was selected as the material of nanoparticles. Experiments were performed for upward- and downward-facing surfaces. Although the CHFs for the downward-facing surface were generally lower than those for the upward-facing surface, the CHFs for the nanoparticle deposited surface were about 1.9 times greater than those for the bare surface in both the configurations. The CHF improvement corresponded well to the reduction of the surface contact angle. During the nucleate boiling in nanofluids, the boiling heat transfer showed peculiar behavior; it was first deteriorated, then improved, and finally approached to an equilibrium state. This observation indicated that the present nanofluid had competing effects to deteriorate and improve the nucleate boiling heat transfer. It was assumed that the wettability and the roughness of the heated surface were influenced by the deposited nanoparticles to cause complex variation of the number of active nucleation sites. During the nucleate boiling of pure water using the downward-facing surface, a sudden increase in the wall temperature was observed stochastically probably due to the accumulation of bubbles beneath the heated surface. Such behavior was not observed when the pure water was replaced by the nanofluid. (author)

  20. Nonequilibrium Green's functions in the study of heat transport of driven nanomechanical systems

    OpenAIRE

    Arrachea, L.; Rizzo, B.

    2013-01-01

    We review a recent theoretical development based on non-equilibrium Green's function formalism to study heat transport in nanomechanical devices modeled by phononic systems of coupled quantum oscillators driven by ac forces and connected to phononic reservoirs. We present the relevant equations to calculate the heat currents flowing along different regions of the setup, as well as the power developed by the time-dependent forces. We also present different strategies to evaluate the Green's fu...

  1. Nonlinear heat and particle transport due to collisional drift waves

    Energy Technology Data Exchange (ETDEWEB)

    Nishi-Kawa, K.I.; Hatori, T.; Terashima, Y.

    1978-07-01

    A nonlinear analysis of collisional drift instability is developed in a slab model based on the two fluid equations, where inhomogeneities in electron and ion temperatures and unperturbed current are included in addition to ion inertia, finite ion gyroradius, and viscosity. A systematic expansion is introduced by taking epsilon=vertical-barkappavertical-barl as a smallness parameter, where kappa is the degree of density gradient and l is the linear scale of the slab along the density gradient. The nonlinear development of the drift wave near marginal stability is studied on the basis of the model equations. A new feature, hard excitation, has been found, which is due to the effects of the nonlinear frequency shift and the electron temperature gradient. The saturation amplitude is calculated, and the expressions for wave-associated particle and heat fluxes are obtained. A comparison of the expressions with the experimental results of a stellerator device is also made.

  2. An Interactive Energy System with Grid, Heating and Transportation Systems

    DEFF Research Database (Denmark)

    Diaz de Cerio Mendaza, Iker

    , flexibility definition and quantification, stochastic impact assessment of LV networks and control of the demand response in LV networks. In a first stage, different residential and non-residential loads are modelled with system analysis purposes. The active loads considered can be categorized as...... of this research work the control of the demand response in LV networks is tackled. The hierarchical structure presented aims to control the operation of heat pumps and plug-in electric vehicles to satisfy technical and commercial aspects of LV grids. This strategy allows system operators to perform their energy......The environmental consciousness and the fact of achieving greater energy independency have led many countries to apply important changes in their energy systems. The intensive renewable energy growth of the last decades represents the most notorious example at this moment. However, this is not only...

  3. Screening for heat transport by groundwater in closed geothermal systems.

    Science.gov (United States)

    Ferguson, Grant

    2015-01-01

    Heat transfer due to groundwater flow can significantly affect closed geothermal systems. Here, a screening method is developed, based on Peclet numbers for these systems and Darcy's law. Conduction-only conditions should not be expected where specific discharges exceed 10(-8)  m/s. Constraints on hydraulic gradients allow for preliminary screening for advection based on rock or soil types. Identification of materials with very low hydraulic conductivity, such as shale and intact igneous and metamorphic rock, allow for analysis with considering conduction only. Variability in known hydraulic conductivity allows for the possibility of advection in most other rocks and soil types. Further screening relies on refinement of estimates of hydraulic gradients and hydraulic conductivity through site investigations and modeling until the presence or absence of conduction can be confirmed. © 2014, National Ground Water Association.

  4. Computational study of heat transport in compositionally disordered binary crystals

    International Nuclear Information System (INIS)

    Lyver, John W.; Blaisten-Barojas, Estela

    2006-01-01

    The thermal conductivity of compositionally disordered binary crystals with atoms interacting through Lennard-Jones potentials has been studied as a function of temperature. The two species in the crystal differ in mass, hard-core atomic diameter, well depth and relative concentration. The isobaric Monte Carlo was used to equilibrate the samples at near-zero pressure. The isoenergy molecular dynamics combined with the Green-Kubo approach was taken to calculate the heat current time-dependent autocorrelation function and determine the lattice thermal conductivity of the sample. The inverse temperature dependence of the lattice thermal conductivity was shown to fail at low temperatures when the atomic diameters of the two species differ. Instead, the thermal conductivity was nearly a constant across temperatures for species with different atomic diameters. Overall, it is shown that there is a dramatic decrease of the lattice thermal conductivity with increasing atomic radii ratio between species and a moderate decrease due to mass disorder

  5. A predictive transport modeling code for ICRF-heated tokamaks

    International Nuclear Information System (INIS)

    Phillips, C.K.; Hwang, D.Q.

    1992-02-01

    In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3. Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5

  6. Heat and mass transport during a groundwater replenishment trial in a highly heterogeneous aquifer

    Science.gov (United States)

    Seibert, Simone; Prommer, Henning; Siade, Adam; Harris, Brett; Trefry, Mike; Martin, Michael

    2014-12-01

    Changes in subsurface temperature distribution resulting from the injection of fluids into aquifers may impact physiochemical and microbial processes as well as basin resource management strategies. We have completed a 2 year field trial in a hydrogeologically and geochemically heterogeneous aquifer below Perth, Western Australia in which highly treated wastewater was injected for large-scale groundwater replenishment. During the trial, chloride and temperature data were collected from conventional monitoring wells and by time-lapse temperature logging. We used a joint inversion of these solute tracer and temperature data to parameterize a numerical flow and multispecies transport model and to analyze the solute and heat propagation characteristics that prevailed during the trial. The simulation results illustrate that while solute transport is largely confined to the most permeable lithological units, heat transport was also affected by heat exchange with lithological units that have a much lower hydraulic conductivity. Heat transfer by heat conduction was found to significantly influence the complex temporal and spatial temperature distribution, especially with growing radial distance and in aquifer sequences with a heterogeneous hydraulic conductivity distribution. We attempted to estimate spatially varying thermal transport parameters during the data inversion to illustrate the anticipated correlations of these parameters with lithological heterogeneities, but estimates could not be uniquely determined on the basis of the collected data.

  7. Superdiffusive Heat Transport in a Class of Deterministic One-dimensional Many-Particle Lorentz Gases

    Science.gov (United States)

    Collet, Pierre; Eckmann, Jean-Pierre; Mejía-Monasterio, Carlos

    2009-07-01

    We study heat transport in a one-dimensional chain of a finite number N of identical cells, coupled at its boundaries to stochastic particle reservoirs. At the center of each cell, tracer particles collide with fixed scatterers, exchanging momentum. In a recent paper (Collet and Eckmann in Commun. Math. Phys. 287:1015, 2009), a spatially continuous version of this model was derived in a scaling regime where the scattering probability of the tracers is γ˜1/ N, corresponding to the Grad limit. A Boltzmann-like equation describing the transport of heat was obtained. In this paper, we show numerically that the Boltzmann description obtained in Collet and Eckmann (Commun. Math. Phys. 287:1015, 2009) is indeed a bona fide limit of the particle model. Furthermore, we study the heat transport of the model when the scattering probability is 1, corresponding to deterministic dynamics. Thought as a lattice model in which particles jump between different scatterers the motion is persistent, with a persistence probability determined by the mass ratio among particles and scatterers, and a waiting time probability distribution with algebraic tails. We find that the heat and particle currents scale slower than 1/ N, implying that this model exhibits anomalous heat and particle transport.

  8. Study of heat transfer and particle transport in Tore Supra and HL-2A tokamaks

    International Nuclear Information System (INIS)

    Song, S.

    2011-12-01

    This thesis reports on experimental studies of heat and particles transport performed on 2 large tokamaks: Tore Supra (based at CEA/Cadarache, France) and HL-2A (based at the Southwestern Institute of Physics, Chengdu, China). The modulated source is the Electron Cyclotron Resonance Heating (ECRH) for the heat pinch and density pump-out studies, while the non-local transport experiments use the Supersonic Molecular Beam Injection (SMBI) as source of modulation. The emphasis is put on the inward heat pinch. In the off-axis ECRH modulation experiments on Tore Supra with low frequency (1 Hz), strong heat inward transport has been observed, in particular for low density. Two transport models have been applied in order to analyze the experimental behavior. The first one is the linear pinch model (LPM) and the second one is an empirical model based on micro-instabilities theory, named Critical Gradient Model (CGM). Good agreement has been found for all harmonics between the experimental data and the simulation using LPM. On the other hand, good agreement has not been achieved using CGM. The density pump-out with large particles and energy losses during ECRH is commonly observed in tokamaks. A new dynamic approach using the modulation technique has been used in HL-2A for analyzing the transient phase of the density pump-out. A correlation between the turbulence increase and the density pump-out has been found. The non-local transport phenomenon, characterized by a fast transient process compared to the normal diffusive response to the perturbation is observed. Both phenomena, i.e., pump-out and non-locality, show as simultaneous variation of density and temperature. This can be an inspiration for the usage of a transport matrix which considers the density and temperature evolution together. Simulations with a simple transport matrix, with non-diagonal terms coupling temperature and density qualitatively reproduce the non-local and pump-out effects qualitatively

  9. Required momentum, heat, and mass transport experiments for liquid-metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Sze, D.K.; Abdou, M.A.

    1986-01-01

    Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

  10. Flexibility analysis of main primary heat transport system : Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Rastogi, S.K.

    1975-01-01

    The paper presents flexibility analysis problem of main primary heat transport system and the approximate analysis that has been made to estimate the loads coming on major equipments. The primary heat transport system for Narora Atomic Power Project is adopting vertical steam generators and pumps equally divided on either side of the reactor with inter-connecting pipes and feeders. Since the system is mainly spring supported with movement of a few points in certain direction defined but no anchorage, it represents a good problem for flexibility analysis which can only be solved in one step by developing a good computer programme. (author)

  11. Design considerations for CRBRP heat transport system piping operating at elevated temperatures

    International Nuclear Information System (INIS)

    Pollono, L.P.; Mello, R.M.

    1979-01-01

    The heat transport system sodium piping for the Clinch River Breeder Reactor Plant (CRBRP) within the reactor containment building must withstand high temperatures for long periods of time. Each phase of the mechanical design process of the piping system is influenced by elevated temperature considerations which include material thermal creep effects, ratchetting caused by rapid temperature transients and stress relaxation, and material degradation effects. The structural design philosophy taken to design the CRBRP piping operating in a high temperature environment is described. The resulting design of the heat transport system piping is presented along with a discussion of special features that resulted from the elevated temperature considerations

  12. Heat transport in low-dimensional materials: A review and perspective

    Directory of Open Access Journals (Sweden)

    Zhiping Xu

    2016-05-01

    Full Text Available Heat transport is a key energetic process in materials and devices. The reduced sample size, low dimension of the problem and the rich spectrum of material imperfections introduce fruitful phenomena at nanoscale. In this review, we summarize recent progresses in the understanding of heat transport process in low-dimensional materials, with focus on the roles of defects, disorder, interfaces, and the quantum-mechanical effect. New physics uncovered from computational simulations, experimental studies, and predictable models will be reviewed, followed by a perspective on open challenges.

  13. Heat transfer analysis of a volumetric solar receiver by coupling the solar radiation transport and internal heat transfer

    International Nuclear Information System (INIS)

    Chen, Xue; Xia, Xin-Lin; Liu, Hua; Li, Yang; Liu, Bo

    2016-01-01

    Highlights: • A model coupling solar radiation transport and internal heat transfer is developed. • Two other treatment approaches for the concentrated solar radiation are compared. • Porous parameters significantly affect the distribution of absorbed solar radiation. • The TBC approach overestimates the solid temperature with a deviation up to 76.4%. • The CIR approach provides acceptable temperature field with deviation less than 3.4%. - Abstract: Volumetric receivers have become a promising technology for the solar thermal conversion. The absorption of concentrated solar radiation and the heat transfer to the working fluid are the two dominant processes. To effectively investigate the thermal performance of receiver, a numerical model coupling the solar radiation transport and the internal heat transfer is presented. Solar radiation transport from the dish concentrator to the interior of receiver is simulated with the Monte Carlo ray tracing method. Combining the distribution of absorbed solar energy in the receiver, the local thermal non-equilibrium model with P1 approximation is used to solve the internal heat transfer. Two other treatment approaches for the concentrated solar radiation are compared. One considers the solar radiation on the front surface of receiver as thermal boundary condition (TBC) and the other as a collimated incident radiation (CIR) beam. The results show that the porosity and mean cell size have a great effect on the distribution of absorbed solar radiation. Compared with the coupling approach, the TBC approach overestimates the solid temperature near the front surface with a deviation up to 76.4%, while the CIR approach provides acceptable temperature field with a deviation less than 3.4%. In addition, the fluid and solid temperatures both decrease as the slope error of concentrator increases.

  14. Evaluated Nuclear Data Library for Transport Calculations at Energies up to 150 MeV

    International Nuclear Information System (INIS)

    Korovin, Yu.A.; Konobeyev, A.Yu.; Pilnov, G.B.; Stankovskiy, A.Yu.

    2005-01-01

    A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The first version of neutron sub-library has been completed and described in the present paper. The library contains nuclear data for transport, heating, and shielding applications for 242 nuclides ranging in atomic number from 8 to 82 in the energy region of primary neutrons from 10-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (Revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The evaluation of emitted particle energy and angular distributions at the energies above 20 MeV was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross sections, elastic cross sections, and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3m or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF/B-VI format using the MF=3/MT=5 and MF=6/MT=5 representations

  15. Turbulent heat transport in two- and three-dimensional temperature fields

    Energy Technology Data Exchange (ETDEWEB)

    Samaraweera, Don Sarath Abesiri [Univ. of California, Berkeley, CA (United States)

    1978-03-01

    A fundamental numerical study of turbulent heat and mass transport processes in two- and three-dimensional convective flows is presented. The model of turbulence employed is the type referred to as a second-order closure. In this scheme transport equations for all nonzero components of the Reynolds stress tensor, for the isotropic dissipation rate of turbulent kinetic energy, for all nonzero scalar flux tensor components and for the mean square scalar fluctuations are solved by a finite difference method along with the mean momentum and mean enthalpy (or concentration) equations. The model used for the stresses was developed earlier. Parallel ideas were utilised in obtaining a model for turbulent heat and mass transfer processes. The study has focused especially on the problem of nonaxisymmetric convective heat and mass transport in pipes, which arises when the boundary conditions are not axisymmetric. The few available experimental data on such situations have indicated anisotropy in effective diffusivities. To expand the available data base an experiment was conducted to obtain heat transfer measurements in strong three-dimensional heating conditions. Numerical procedures especially suitable for incorporation of second-order turbulent closure models have been developed. The effect of circumferential conduction in the tube material, which is influential in the asymmetric heating data currently available, was accounted for directly by extending the finite difference calculations into the pipe wall. The principal goal of predicting three-dimensional scalar transfer has been achieved.

  16. Analytical and experimental investigations of the passive heat transport in HTRs under severe accident conditions

    International Nuclear Information System (INIS)

    Rehm, W.; Barthels, H.; Jahn, W.; Cleveland, J.C.; Ishihara, M.

    1992-01-01

    Thermodynamic accident analyses have been performed with computer simulation models to investigate core heatup sequences, sensitivity analyses, power variations, anticipated transients without scram, and core displacement considerations for probabilistic safety analyses (PSA) of small gas-cooled high-temperature reactors (e.g. HTR-Module). In worst case considerations where not only a loss of the active heat removal system is assumed but also a loss of the vessel cooling system, the heat would be transported into the surrounding concrete structure. In such a case the concrete would act as a natural long-term intermediate heat storage dissipating the heat through the concrete surface. Large scale and reactor safety experiments have been performed to investigate passive heat transport mechanisms -- which can cooldown a HIR core during severe accident conditions -- for validation basis of computer simulation codes used for accident analyses. In general, the comparisons of experimental and analytical results with computer calculations of the heat transport codes are in good agreement

  17. Unsteady heat conduction in the soil layers above underground repository for spent nuclear fuel

    Science.gov (United States)

    Talukder, N. K.

    Due to radioactivity of spent nuclear fuel, a repository is expected to act as a heat source of exponentially decreasing intensity over hundreds of years. In case of underground emplacement of such a heat source, the temperature changes in the soil layers surrounding the heat source may have important implications such as evaporation of the water contained in the soil and its subsequent condensation. Assuming a uniformly distributed power generation over a horizontal, relatively thin, circular zone of several thousand meters diameter located several hundred meters below the ground surface, the temperature variations along the vertical centerline of the heating zone was estimated analytically under simplifying assumptions. Unsteady one-dimensional heat conduction in a semi-infinite solid with an exponentially decreasing heat flux at the proximal end was considered. The corresponding solution of the Fourier equation for heat conduction contains Error Functions of complex arguments. The evaluation of the Error Functions for discrete space and time parameters was performed applying analytical procedures and using standard tables. The results show temperature distributions in the soil at various time points over thousands of years after underground emplacement of spent nuclear fuel.

  18. Design and application for a high-temperature nuclear heat source

    International Nuclear Information System (INIS)

    Quade, R.N.

    1980-01-01

    Recent actions by OPEC have sharply increased interest in the United States in synfuels, with coal being the logical choice for the carbon source. Two coal liquefaction processes, direct and indirect, have been examined. Each can produce about 50% more output when coupled to an HTGR for process heat. The nuclear reactor designed for process heat has a power output of 842MW(t), a core outlet temperature of 950 0 C (1742 0 F), and an intermediate helium loop to separate the heat source from the process heat exchangers. Steam-methane reforming is the reference process. As part of the development of a nuclear process heat system, a computer code, Process Heat Reactor Evaluation and Design, is being developed. This code models both the reactor plant and a steam reforming plant. When complete, the program will have the capability to calculate an overall mass and heat balance, size the plant components, and estimate the plant cost for a wide variety of independent variables. (author)

  19. Heat pulse analysis in JET and relation to local energy transport models

    International Nuclear Information System (INIS)

    Haas, J.C.M. de; Lopes Cardozo, N.J.; Han, W.; Sack, C.; Taroni, A.

    1989-01-01

    The evolution of a perturbation T e of the electron temperature depends on the linearised expression of the heat flux q e and may be not simply related to the local value of the electron heat conductivity χ e . It is possible that local heat transport models predicting similar temperature profiles and global energy confinement properties, imply a different propagation of heat pulses. We investigate here this possibility for the case of two models developed at JET. We also present results obtained at JET on a set of discharges covering the range of currents from 2 to 5 MA. Only L-modes, limiter discharges are considered here. Experimental results on the scaling of χ HP , the value of χ e related to heat pulse propagation, are compared with those of χ HP derived from the models. (author) 7 refs., 2 figs., 2 tabs

  20. Modelling of Temperature Profiles and Transport Scaling in Auxiliary Heated Tokamaks

    DEFF Research Database (Denmark)

    Callen, J.D.; Christiansen, J.P.; Cordey, J.G.

    1987-01-01

    in detail: (i) a heat pinch or excess temperature gradient model with constant coefficients; and (ii) a non-linear heat diffusion coefficient (χ) model. Both models predict weak (lesssim20%) temperature profile responses to physically relevant changes in the heat deposition profile – primarily because...... that result from the models clarify why temperature profiles in many tokamaks are often characterized as exhibiting a high degree of 'profile consistency'. Global transport scaling laws are also derived from the two models. The non-linear model with χ ∝ dT/dr produces a non-linear energy confinement time (L......-mode) scaling with input power, . The constant heat pinch or excess temperature gradient model leads to the offset linear law for the total stored energy W with Pin, W = τinc Pin + W(0), which describes JET auxiliary heating data quite well. It also provides definitions for the incremental energy confinement...

  1. The entropy problem of the decentralized solar and nuclear heat generation

    International Nuclear Information System (INIS)

    Seifritz, W.

    1984-01-01

    Parallel to the energy fluxes the entropy fluxes of decentralized hot-water systems based on solar collectors coupled with an electrical auxiliary heating installation are also deduced. As an important result the fact emerges that this kind of solar energy has to remain very restricted, not only for quantitative-energetic reasons, but also for entropy ones, and that a solar hot-water system will always have to rely on an energy system of low entropy. In contrast to this, the provision of heat for space heating sector with the help of the 'nuclear short-distance concept', which practically does not need any external energy, is not subject to these restrictions. This concept is introduced briefly, as well as the heat prices which presumably can be achieved by it. Concluding comments summarize the reasons once again that speak against the installation of a decentralized solar heat supply system. (orig.) [de

  2. Modification of the finite element heat and mass transfer code (FEHM) to model multicomponent reactive transport

    International Nuclear Information System (INIS)

    Viswanathan, H.S.

    1996-08-01

    The finite element code FEHMN, developed by scientists at Los Alamos National Laboratory (LANL), is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developing hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent Kd model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The new chemical capabilities of FEHMN are illustrated by using Los Alamos National Laboratory's site scale model of Yucca Mountain to model two-dimensional, vadose zone 14 C transport. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also prove that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies

  3. Refrigerant Performance Evaluation Including Effects of Transport Properties and Optimized Heat Exchangers.

    Science.gov (United States)

    Brignoli, Riccardo; Brown, J Steven; Skye, H; Domanski, Piotr A

    2017-08-01

    Preliminary refrigerant screenings typically rely on using cycle simulation models involving thermodynamic properties alone. This approach has two shortcomings. First, it neglects transport properties, whose influence on system performance is particularly strong through their impact on the performance of the heat exchangers. Second, the refrigerant temperatures in the evaporator and condenser are specified as input, while real-life equipment operates at imposed heat sink and heat source temperatures; the temperatures in the evaporator and condensers are established based on overall heat transfer resistances of these heat exchangers and the balance of the system. The paper discusses a simulation methodology and model that addresses the above shortcomings. This model simulates the thermodynamic cycle operating at specified heat sink and heat source temperature profiles, and includes the ability to account for the effects of thermophysical properties and refrigerant mass flux on refrigerant heat transfer and pressure drop in the air-to-refrigerant evaporator and condenser. Additionally, the model can optimize the refrigerant mass flux in the heat exchangers to maximize the Coefficient of Performance. The new model is validated with experimental data and its predictions are contrasted to those of a model based on thermodynamic properties alone.

  4. Heat science and transport phenomena in fuel cells; Thermique et phenomenes de transport dans les piles a combustible

    Energy Technology Data Exchange (ETDEWEB)

    Liberatore, P.M.; Boillot, M. [Laboratoire des Sciences du Genie Chimique de Nancy, 54 - Vandoeuvre-les-Nancy (France); Bonnet, C.; Didieerjean, S.; Lapicque, F.; Deseure, J.; Lottin, O.; Maillet, D.; Oseen-Senda, J. [Laboratoire d' Energetique et de Mecanique Theorique et Appliquee, 54 - Vandoeuvre Les Nancy (France); Alexandre, A. [Laboratoire d' Etudes Thermiques, ENSMA, 86 Poitiers (France); Topin, F.; Occelli, R.; Daurelle, J.V. [IUSTI / Polytech' Marseille, Institut universitaire des Systemes Thermiques Industriels Ecole, 13 - Marseille (France); Pauchet, J.; Feidt, M. [CEA Grenoble, Groupement pour la recherche sur les echangeurs thermiques (Greth), 38 (France); Voarino, C. [CEA Centre d' Etudes du Ripault, 37 - Tours (France); Morel, B.; Laurentin, J.; Bultel, Y.; Lefebvre-Joud, F. [CEA Grenoble, LEPMI, 38 (France); Auvity, B.; Lasbet, Y.; Castelain, C.; Peerohossaini, H. [Ecole Centrale de Nantes, Laboratoire de Thermocinetique de Nantes (LTN), 44 - Nantes (France)

    2005-07-01

    In this work are gathered the transparencies of the lectures presented at the conference 'heat science and transport phenomena in fuel cells'. The different lectures have dealt with 1)the gas distribution in the bipolar plates of a fuel cell: experimental studies and computerized simulations 2)two-phase heat distributors in the PEMFC 3)a numerical study of the flow properties of the backing layers on the transfers in a PEMFC 4)modelling of the heat and mass transfers in a PEMFC 5)two-phase cooling of the PEMFC with pentane 6)stationary thermodynamic model of the SOFC in the GECOPAC system 7)modelling of the internal reforming at the anode of the SOFC 8)towards a new thermal design of the PEMFC bipolar plates. (O.M.)

  5. Nuclear energy as an alternative heat energy source for oil refinery

    International Nuclear Information System (INIS)

    Sunardi; Djati H Salimy; Edwaren Liun; Sahala M Lumbanraja

    2012-01-01

    The study of high temperature nuclear heat application for oil refinery has been carried out. The goal of the study is to understand the characteristic and possibility of high temperature nuclear heat application for operation of oil refinery. In this study, the oil refinery plant with the capacity crude oil processing of 126 MBSD is used as a reference. It is assumed that high temperature of nuclear energy will supply steam and electricity to the plant, while the high temperature processes sill fossil fuel firing utilizing. Nuclear reactor that used in this study is small high temperature nuclear reactor HTR-PM250. Energy balance calculations indicate that thermal energie from nuclear reactor with thermal power of 250 MWt can be distributed as follow: 41,23 MWt to produce steam 1 (385°C, 40 kg/cm2, 55,6 ton/hr), 101,47MWt to produce steam 2 (360°C, 15 kg/cm2, 131,1 ton/jam), and 60 MWt to produce 24 MWe electricity. The rest, about 22,3 MWt is converted to electricity about 8,93 MWe, send to the public grid. Utilization nuclear energy to substitute steam and electricity production for oil refinery with capacity of 126 MBSD crude oil, give reduction of fossil fuel burning about 64,8 thousand ton/yr, equivalent to reduce CO2 emission about 182;4 thousand ton/yr. (author)

  6. The role of monsoon-like zonally asymmetric heating in interhemispheric transport

    Science.gov (United States)

    Chen, Gang; Orbe, Clara; Waugh, Darryn

    2017-03-01

    While the importance of the seasonal migration of the zonally averaged Hadley circulation on interhemispheric transport of trace gases has been recognized, few studies have examined the role of the zonally asymmetric monsoonal circulation. This study investigates the role of monsoon-like zonally asymmetric heating on interhemispheric transport using a dry atmospheric model that is forced by idealized Newtonian relaxation to a prescribed radiative equilibrium temperature. When only the seasonal cycle of zonally symmetric heating is considered, the mean age of air in the Southern Hemisphere since last contact with the Northern Hemisphere midlatitude boundary layer is much larger than the observations. The introduction of monsoon-like zonally asymmetric heating not only reduces the mean age of tropospheric air to more realistic values but also produces an upper tropospheric cross-equatorial transport pathway in boreal summer that resembles the transport pathway simulated in the NASA Global Modeling Initiative Chemistry Transport Model driven with Modern-Era Retrospective Analysis for Research and Applications meteorological fields. These results highlight that the monsoon-induced eddy circulation plays an important role in the interhemispheric transport of long-lived chemical constituents.

  7. Transport of Heat and Charge in Electromagnetic Metrology Based on Nonequilibrium Statistical Mechanics

    Directory of Open Access Journals (Sweden)

    James Baker-Jarvis

    2009-11-01

    Full Text Available Current research is probing transport on ever smaller scales. Modeling of the electromagnetic interaction with nanoparticles or small collections of dipoles and its associated energy transport and nonequilibrium characteristics requires a detailed understanding of transport properties. The goal of this paper is to use a nonequilibrium statistical-mechanical method to obtain exact time-correlation functions, fluctuation-dissipation theorems (FD, heat and charge transport, and associated transport expressions under electromagnetic driving. We extend the time-symmetric Robertson statistical-mechanical theory to study the exact time evolution of relevant variables and entropy rate in the electromagnetic interaction with materials. In this exact statistical-mechanical theory, a generalized canonical density is used to define an entropy in terms of a set of relevant variables and associated Lagrange multipliers. Then the entropy production rate are defined through the relevant variables. The influence of the nonrelevant variables enter the equations through the projection-like operator and thereby influences the entropy. We present applications to the response functions for the electrical and thermal conductivity, specific heat, generalized temperature, Boltzmann’s constant, and noise. The analysis can be performed either classically or quantum-mechanically, and there are only a few modifications in transferring between the approaches. As an application we study the energy, generalized temperature, and charge transport equations that are valid in nonequilibrium and relate it to heat flow and temperature relations in equilibrium states.

  8. 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.

  9. Moment approach to neoclassical flows, currents and transport in auxiliary heated tokamaks

    International Nuclear Information System (INIS)

    Kim, Yil Bong.

    1988-02-01

    The moment approach is utilized to derive the full complement of neoclassical transport processes in auxiliary heated tokamaks. The effects of auxiliary heating [neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH)] considered arise from the collisional interaction between the background plasma species and the fast-ion-tail species. From a known fast ion distribution function we evaluate the parallel (to the magnetic field) momentum and heat flow inputs to the background plasma. Then, through the momentum and heat flow balance equations, we can determine the induced parallel flows (and current) and radial transpot fluxes in ''equilibrium'' (on the time scale much longer than the collisional relaxation time, i.e., t >> 1ν/sub ii/). In addition to the fast-ion-induced current, the total neoclassical current includes the boostap current, which is driven by the pressure and temperature gradients, the Pfirsch-Schlueter current which is required for charge neutrality, and the neoclassical (including trapped particle effects) Spitzer current due to the parallel electric field. The radial transport fluxes also include off-diagonal compnents in the transport matrix which correspond to the Ware (neoclassical) pinch due to the inductive applied electric field an the fast-ion-induced radial fluxes, in addition to the usual pressure- and temperature-gradient-driven fluxes (particle diffusion and heat conduction). Once the tranport coefficient are completely determined, the radial fluxes and the heat fluxes can be substituted into the density and energy evolution equations to provide a complete description of ''equilibrium'' (δδt << ν/sub ii/) neoclassical transport processes in a plasma. 47 refs., 14 figs

  10. The role of individual cyclones for atmospheric latent and sensible heat transport into the European Arctic

    Science.gov (United States)

    Sodemann, H.; Stohl, A.

    2010-12-01

    The bulk of the atmospheric latent heat transport induced by extratropical cyclones is organized in the warm conveyor belt, also known as atmospheric rivers. In order to enhance the process understanding of atmospheric sensible and latent heat transport with these structures into the European Arctic, the magnitude and variability of the energy flux from individual cyclones in this region was studied. We applied a moisture source tracking algorithm embedded in the limited-area numerical weather prediction model (NWP) Climate High-Resolution Model (CHRM) to trace the evaporation sources and transport of water vapour from different latitude bands of the North Atlantic Ocean. September 2002 and December 2006 were chosen as initial analysis periods, since a particularly large number of cyclones (including former hurricanes) traveled within the North Atlantic storm track during these months. The main findings are that latent heat (LH) from more southerly source regions is transported at higher altitudes. Stronger storms draw latent heat from a larger area (further south), and the ensuing precipitation will hence on average originate from further south as well. Most long-range transport of LH occurs in the cold frontal bands. Individual cyclones are the main source of sub-monthly LH flux variability, and can cause up to 4-sigma variation of the mean flux. LH flux is almost permanently net positive (northward), unlike for sensible heat (SH) and other energy fluxes. Most LH that is "permanently" transferred to north of 60°N in the Atlantic storm track originates from directly south of that latitude, implying on average short atmospheric moisture lifetimes, and hence a fast energy turnover. We compare these findings to results from a Lagrangian moisture tracking method based on the FLEXPART model. Remarks with regard to differences in the transport conditions of latent head in such structures along the North American West Coast and the Norwegian West Coast will be made.

  11. Progress and safety aspects in process heat utilization from nuclear systems

    International Nuclear Information System (INIS)

    Barnert, H.

    1995-01-01

    Report about the Status and the Progress in the Various Programs and Projects in the Federal Republic of Germany in Process Heat Utilization from the High Temperature Reactor and on Recent Changes of the Atomic Law in the Federal Republic of Germany with Big Influence on the Safety of Nuclear Energy Technology. (author)

  12. Heating- and growing-degree days at Chalk River Nuclear Laboratories, 1976-1980

    International Nuclear Information System (INIS)

    Jay, P.C.; Wildsmith, D.P.

    1981-05-01

    An update of the report, Heating- and Growing-Degree-Days at Chalk River Nuclear Laboratories (AECL-5547) is presented along with various other meteorological variables which were not included in the previous publication. Also included, and shown in graph form, are the monthly degree-day frequencies. (author)

  13. An overview of heat exchanger technology in the Canadian nuclear program

    International Nuclear Information System (INIS)

    Carlucci, L.N.; Dalrymple, D.G.; Ko, P.L.; Pathania, R.; Pettigrew, M.I.; Scott, D.A.

    1981-06-01

    This paper provides an overview of the Canadian approach to the reliability and serviceability of heat exchange equipment used in nuclear power stations and heavy water plants. Current work in vibration and fretting predictions, thermal-hydraulic analyses, and corrosion research is described. Procedures developed for in-service inspection, in situ tube replacment and chemical cleaning of corrosion products are also outlined

  14. The application of high temperature nuclear heat for urea fertilizer plant

    International Nuclear Information System (INIS)

    Djati H Salimy

    2012-01-01

    The study of high temperature nuclear heat application for urea fertilizer plant has been carried out. The goal of the study is to understand the characteristic and possibility of high temperature nuclear heat application for urea fertilizer plant. In this study, the urea fertilizer plant with the capacity production of 1725 ton per day is used as a reference. Energy balance calculating indicate that thermal energie from high temperature nuclear reactor with thermal power of 600 MWt, can be distributed as follow: 274,3 MWt can be utilized as process heat for natural gas steam reforming, while about 101,44 MWt for process steam and electricity production to support the process. Total energy required for operating the plant is 375,74 MWt. The rest, about 101,44 MWt is converted to electricity about 54,2 MWe, send to the public grid. In conventional process of urea fertilizer plant, the demand of natural gas is used for raw material, heat energy source for chemical reaction, and utility (process steam and electricity), while for nuclear application the need of natural gas only for raw material. At conventional process, from total demand of natural gas about 21.25 million MMBTU per year, only about 8.95 million MMBTU used as raw material. n About 12,3 million MMBTU (-60%) natural gas can be saved that mean reduce Co2 emission about 718,192.42 ton/year. (author)

  15. The Influence of Heat Flux Boundary Heterogeneity on Heat Transport in Earth's Core

    Science.gov (United States)

    Davies, C. J.; Mound, J. E.

    2017-12-01

    Rotating convection in planetary systems can be subjected to large lateral variations in heat flux from above; for example, due to the interaction between the metallic cores of terrestrial planets and their overlying silicate mantles. The boundary anomalies can significantly reorganise the pattern of convection and influence global diagnostics such as the Nusselt number. We have conducted a suite of numerical simulations of rotating convection in a spherical shell geometry comparing convection with homogeneous boundary conditions to that with two patterns of heat flux variation at the outer boundary: one hemispheric pattern, and one derived from seismic tomographic imaging of Earth's lower mantle. We consider Ekman numbers down to 10-6 and flux-based Rayleigh numbers up to 800 times critical. The heterogeneous boundary conditions tend to increase the Nusselt number relative to the equivalent homogeneous case by altering both the flow and temperature fields, particularly near the top of the convecting region. The enhancement in Nusselt number tends to increase as the amplitude and wavelength of the boundary heterogeneity is increased and as the system becomes more supercritical. In our suite of models, the increase in Nusselt number can be as large as 25%. The slope of the Nusselt-Rayleigh scaling also changes when boundary heterogeneity is included, which has implications when extrapolating to planetary conditions. Additionally, regions of effective thermal stratification can develop when strongly heterogeneous heat flux conditions are applied at the outer boundary.

  16. Nuclear data production, calculation and measurement: a global overview of the gamma heating issue

    Directory of Open Access Journals (Sweden)

    Gueton O.

    2013-03-01

    Full Text Available The gamma heating evaluation in different materials found in current and future generations of nuclear reactor (EPRTM, GENIV, MTR-JHR, is becoming an important issue especially for the design of many devices (control rod, heavy reflector, in-core & out-core experiments…. This paper deals with the works started since 2009 in the Reactor Studies Department of CEA Cadarache in ordre to answer to several problematic which have been identified as well for nuclear data production and calculation as for experimental measurement methods. The selected subjects are: Development of a Monte Carlo code (FIFRELIN to simulate the prompt fission gamma emission which represents the major part of the gamma heating production inside the core Production and qualification of new evaluations of nuclear data especially for radiative capture and inelastic neutron scattering which are the main sources of gamma heating out-core Development and qualification of a recommended method for the total gamma heating calculation using the Monte Carlo simulation code TRIPOLI-4 Development, test and qualification of new devices dedicated to the in-core gamma heating measurement as well in MTR-JHR as in zero power facilities (EOLE-MINERVE of CEA, Cadarache to increase the experimental measurement accuracy.

  17. Heat transfer in the decay pool of irradiated nuclear fuel

    International Nuclear Information System (INIS)

    Lorenzini, E.; Spiga, M.; Zerbini, D.

    1987-01-01

    Present work aims to analyze mono-dimensional, time dependent trends of the water temperature, assuming that there are no auxiliary cooling systems; this can simulate one of the worst hypothetical accidents which may occur in a nuclear plant. In fact the cooling of the spent fuel is provided only by natural convection; in this situation it is an interesting matter to valuate the time evolution of the water temperature and its maximum value. The accuracy of this method requires a very large computer time and makes it rather heavy and expensive. In this paper, a method to perform the mono-dimensional transient thermal hydraulic analysis of the storage pool is proposed. The pool is subdivided in horizontal isothermal segments; for every segment an enthalpy balance equation is deduced, taking account of a vertical mixing flow rate between adjacent segments, due to the convective fluid streams up and down

  18. Summary report on transportation of nuclear fuel materials in Japan : transportation infrastructure, threats identified in open literature, and physical protection regulations.

    Energy Technology Data Exchange (ETDEWEB)

    Cochran, John Russell; Ouchi, Yuichiro (Japan Atomic Energy Agency, Japan); Furaus, James Phillip; Marincel, Michelle K.

    2008-03-01

    This report summarizes the results of three detailed studies of the physical protection systems for the protection of nuclear materials transport in Japan, with an emphasis on the transportation of mixed oxide fuel materials1. The Japanese infrastructure for transporting nuclear fuel materials is addressed in the first section. The second section of this report presents a summary of baseline data from the open literature on the threats of sabotage and theft during the transport of nuclear fuel materials in Japan. The third section summarizes a review of current International Atomic Energy Agency, Japanese and United States guidelines and regulations concerning the physical protection for the transportation of nuclear fuel materials.

  19. Study of Nuclear Decay Data Contribution to Uncertainties in Heat Load Estimations for Spent Fuel Pools

    Science.gov (United States)

    Ferroukhi, H.; Leray, O.; Hursin, M.; Vasiliev, A.; Perret, G.; Pautz, A.

    2014-04-01

    At the Paul Scherrer Institut (PSI), a methodology for nuclear data uncertainty propagation in CASMO-5M (C5M) assembly calculations is under development. This paper presents a preliminary application of this methodology to C5M decay heat calculations. Applying a stochastic sampling method, nuclear decay data uncertainties are first propagated for the cooling phase only. Thereafter, the uncertainty propagation is enlarged to gradually account for cross-section as well as fission yield uncertainties during the depletion phase. On that basis, assembly heat load uncertainties as well as total uncertainty for the entire pool are quantified for cooling times up to one year. The relative contributions from the various types of nuclear data uncertainties are in this context also estimated.

  20. A study of a small nuclear power plant system for district heating

    International Nuclear Information System (INIS)

    Imamura, Mitsuru; Sato, Kotaro; Narabayashi, Tadashi; Shimazu, Yoichiro; Tsuji, Masashi

    2009-01-01

    We have studied nuclear power plant for district heating. Already some towns and villages in Hokkaido have requested small reactor for district heating. Using existing technology allows us to shorten development period and to keep a lid on development cost. We decided to develop new reactor based on 'MUTSU' reactor technology because 'MUTSU' had already proved its safety. And this reactor was boron free reactor. It allows plant system to reduce the chemical control system. And moderator temperature coefficient is deeply negative. It means to improve its operability and leads to dependability enhancement. We calculated burn-up calculation of erbium addition fuel. In the result, the core life became about 10 years. And we adapt the cassette type refueling during outagein in order to maintain nonproliferation. In the district heating system, a double heat exchanger system enables to response to load change in season. To obtain the acceptance of public, this system has a leak prevention system of radioactive materials to public. And road heating system of low grade heat utilization from turbine condenser leads to improve the heat utilization efficiency. We carried out performance evaluation test of district heating pipeline. Then the heat loss of pipeline is estimated at about 0.440degC/km. This result meets general condition, which is about 1degC/km. This small plant has passive safety system. It is natural cooling of containment vessel. In case of loss of coolant accident, decay heat can remove by natural convection air cooling after 6 hours. Decay heat within 6 hours can remove by evaporative heat transfer of pool on containment vessel. (author)

  1. Numerical modeling of coupled water flow and heat transport in soil and snow

    Science.gov (United States)

    Thijs J. Kelleners; Jeremy Koonce; Rose Shillito; Jelle Dijkema; Markus Berli; Michael H. Young; John M. Frank; William Massman

    2016-01-01

    A one-dimensional vertical numerical model for coupled water flow and heat transport in soil and snow was modified to include all three phases of water: vapor, liquid, and ice. The top boundary condition in the model is driven by incoming precipitation and the surface energy balance. The model was applied to three different terrestrial systems: A warm desert bare...

  2. Stochastic Impact Assessment of the Heating and Transportation Systems Electrification on LV grids

    DEFF Research Database (Denmark)

    Mendaza, Iker Diaz de Cerio; Bak-Jensen, Birgitte; Chen, Zhe

    2014-01-01

    According to the new energy policy agreements, a conceptual and technological re-structuration of the Danish energy sector is expected. One of the key points for its successful implementation is the partial electrification of the heating and transportation systems. This fact, which reflects...

  3. Study of heat transport in structured soil under grass cover. Dual-continuum approach

    Czech Academy of Sciences Publication Activity Database

    Votrubová, J.; Dohnal, M.; Tesař, Miroslav; Vogel, T.

    2011-01-01

    Roč. 13, - (2011), s. 7414 ISSN 1607-7962. [European Geosciences Union General Assembly 2011. 03.04.2011-08.04.2011, Vienna] R&D Projects: GA ČR GA205/08/1174 Institutional research plan: CEZ:AV0Z20600510 Keywords : water and heat transport * model S1D * Sumava Mts. Subject RIV: DA - Hydrology ; Limnology

  4. Measurements of Combined Axial Mass and Heat Transport in He II.

    Science.gov (United States)

    Johnson, Warren W.; Jones, Michael C.

    An experiment was performed that allowed measurements of both axial mass and heat transport of He-II (the superfluid phase of helium 4) in a long tube. The apparatus allowed the pressure difference and the temperature difference across the flow tube to each be independently adjusted, and the resulting steady-state values of net fluid velocity and…

  5. Thermal transport in low dimensions from statistical physics to nanoscale heat transfer

    CERN Document Server

    2016-01-01

    Understanding non-equilibrium properties of classical and quantum many-particle systems is one of the goals of contemporary statistical mechanics. Besides its own interest for the theoretical foundations of irreversible thermodynamics(e.g. of the Fourier's law of heat conduction), this topic is also relevant to develop innovative ideas for nanoscale thermal management with possible future applications to nanotechnologies and effective energetic resources. The first part of the volume (Chapters 1-6) describes the basic models, the phenomenology and the various theoretical approaches to understand heat transport in low-dimensional lattices (1D e 2D). The methods described will include equilibrium and nonequilibrium molecular dynamics simulations, hydrodynamic and kinetic approaches and the solution of stochastic models. The second part (Chapters 7-10) deals with applications to nano and microscale heat transfer, as for instance phononic transport in carbon-based nanomaterials, including the prominent case of na...

  6. Heat transport in the quasi-single-helicity islands of EXTRAP T2R

    Science.gov (United States)

    Frassinetti, L.; Brunsell, P. R.; Drake, J.

    2009-03-01

    The heat transport inside the magnetic island generated in a quasi-single-helicity regime of a reversed-field pinch device is studied by using a numerical code that simulates the electron temperature and the soft x-ray emissivity. The heat diffusivity χe inside the island is determined by matching the simulated signals with the experimental ones. Inside the island, χe turns out to be from one to two orders of magnitude lower than the diffusivity in the surrounding plasma, where the magnetic field is stochastic. Furthermore, the heat transport properties inside the island are studied in correlation with the plasma current and with the amplitude of the magnetic fluctuations.

  7. Physical protection in the transport of nuclear materials (Legal aspects of the domestic system)

    International Nuclear Information System (INIS)

    Novais, F.J.G.

    1978-04-01

    A study of the physical protection system is made. Emphasis is given to some considerations in the nuclear material transport area, mainly the details of the domestic system, from a juridic pont of view. (Author) [pt

  8. Multiple nucleon transfer in damped nuclear collisions. [Lectures, mass charge, and linear and angular momentum transport

    Energy Technology Data Exchange (ETDEWEB)

    Randrup, J.

    1979-07-01

    This lecture discusses a theory for the transport of mass, charge, linear, and angular momentum and energy in damped nuclear collisions, as induced by multiple transfer of individual nucleons. 11 references.

  9. The Role of Ocean and Atmospheric Heat Transport in the Arctic Amplification

    Science.gov (United States)

    Vargas Martes, R. M.; Kwon, Y. O.; Furey, H. H.

    2017-12-01

    Observational data and climate model projections have suggested that the Arctic region is warming around twice faster than the rest of the globe, which has been referred as the Arctic Amplification (AA). While the local feedbacks, e.g. sea ice-albedo feedback, are often suggested as the primary driver of AA by previous studies, the role of meridional heat transport by ocean and atmosphere is less clear. This study uses the Community Earth System Model version 1 Large Ensemble simulation (CESM1-LE) to seek deeper understanding of the role meridional oceanic and atmospheric heat transports play in AA. The simulation consists of 40 ensemble members with the same physics and external forcing using a single fully coupled climate model. Each ensemble member spans two time periods; the historical period from 1920 to 2005 using the Coupled Model Intercomparison Project Phase 5 (CMIP5) historical forcing and the future period from 2006 to 2100 using the CMIP5 Representative Concentration Pathways 8.5 (RCP8.5) scenario. Each of the ensemble members are initialized with slightly different air temperatures. As the CESM1-LE uses a single model unlike the CMIP5 multi-model ensemble, the internal variability and the externally forced components can be separated more clearly. The projections are calculated by comparing the period 2081-2100 relative to the time period 2001-2020. The CESM1-LE projects an AA of 2.5-2.8 times faster than the global average, which is within the range of those from the CMIP5 multi-model ensemble. However, the spread of AA from the CESM1-LE, which is attributed to the internal variability, is 2-3 times smaller than that of the CMIP5 ensemble, which may also include the inter-model differences. CESM1LE projects a decrease in the atmospheric heat transport into the Arctic and an increase in the oceanic heat transport. The atmospheric heat transport is further decomposed into moisture transport and dry static energy transport. Also, the oceanic heat

  10. Effect of coherence of nonthermal reservoirs on heat transport in a microscopic collision model

    Science.gov (United States)

    Li, Lei; Zou, Jian; Li, Hai; Xu, Bao-Ming; Wang, Yuan-Mei; Shao, Bin

    2018-02-01

    We investigate the heat transport between two nonthermal reservoirs based on a microscopic collision model. We consider a bipartite system consisting of two identical subsystems, and each subsystem interacts with its own local reservoir, which consists of a large collection of initially uncorrelated ancillas. Then a heat transport is formed between two reservoirs by a sequence of pairwise collisions (intersubsystem and subsystem-local reservoir). In this paper we consider two kinds of the reservoir's initial states: the thermal state and the state with coherence whose diagonal elements are the same as that of the thermal state and the off-diagonal elements are nonzero. In this way, we define the effective temperature of the reservoir with coherence according to its diagonal elements. We find that for two reservoirs having coherence the direction of the steady current of heat is different for different phase differences between the two initial states of two reservoirs, especially the heat can transfer from the "cold reservoir" to the "hot reservoir" in the steady regime for particular phase difference. In the limit of the effective temperature difference between the two reservoirs Δ T →0 , for most of the phase differences, the steady heat current increases with the increase of effective temperature until it reaches the high effective temperature limit, while for the thermal state or particular phase difference the steady heat current decreases with the increase of temperature at high temperatures, and in this case the conductance can be obtained.

  11. Analysis of photon transport in biological tissue and the subsequent heating effects

    International Nuclear Information System (INIS)

    Fadhali, M.M.A.

    2015-01-01

    Analysis of laser interaction with matter revealed the possibilities of many industrial and therapeutic applications. This research article discusses the theoretical aspects of laser beam interaction with biological tissues. It introduces the numerical analysis of photon distribution and transport in the tissue and its bio-thermal heating effects. The Monte Carlo method has been applied to simulate the variation of photon distribution and photon fluence with the radial distance from the point of interaction as well as laser powers and tissue thickness. For a specific wavelength, the variation of diffuse reflectance with the absorption coefficient was depicted for different values of the anisotropy factor. It has also been used to simulate the bio-heat transfer to obtain the temperature variation with the heating depth. On the other hand, finite difference method (FDM) has been applied to simulate the heating effect resulted from the incident laser beam on the tissue based on Penne's bio-heat equation combined with the obtained photon distribution and transport parameters from the MC method. The heating effect of the laser beam and hence the occurred thermal damage in the tissue was depicted. A linear relationship between the temperature and the rate of thermal damage has been manifested. This result can be used as a threshold reference for various medical applications of lasers. (authors)

  12. Hydrous mineral dehydration around heat-generating nuclear waste in bedded salt formations.

    Science.gov (United States)

    Jordan, Amy B; Boukhalfa, Hakim; Caporuscio, Florie A; Robinson, Bruce A; Stauffer, Philip H

    2015-06-02

    Heat-generating nuclear waste disposal in bedded salt during the first two years after waste emplacement is explored using numerical simulations tied to experiments of hydrous mineral dehydration. Heating impure salt samples to temperatures of 265 °C can release over 20% by mass of hydrous minerals as water. Three steps in a series of dehydration reactions are measured (65, 110, and 265 °C), and water loss associated with each step is averaged from experimental data into a water source model. Simulations using this dehydration model are used to predict temperature, moisture, and porosity after heating by 750-W waste canisters, assuming hydrous mineral mass fractions from 0 to 10%. The formation of a three-phase heat pipe (with counter-circulation of vapor and brine) occurs as water vapor is driven away from the heat source, condenses, and flows back toward the heat source, leading to changes in porosity, permeability, temperature, saturation, and thermal conductivity of the backfill salt surrounding the waste canisters. Heat pipe formation depends on temperature, moisture availability, and mobility. In certain cases, dehydration of hydrous minerals provides sufficient extra moisture to push the system into a sustained heat pipe, where simulations neglecting this process do not.

  13. Selection of heat disposal methods for a Hanford Nuclear Energy Center

    International Nuclear Information System (INIS)

    Young, J.R.; Kannberg, L.D.; Ramsdell, J.V.; Rickard, W.H.; Watson, D.G.

    1976-06-01

    Selection of the best method for disposal of the waste heat from a large power generation center requires a comprehensive comparison of the costs and environmental effects. The objective is to identify the heat dissipation method with the minimum total economic and environmental cost. A 20 reactor HNEC will dissipate about 50,000 MWt of waste heat; a 40 reactor HNEC would release about 100,000 MWt. This is a much larger discharge of heat than has occurred from other concentrated industrial facilities and consequently a special analysis is required to determine the permissibility of such a large heat disposal and the best methods of disposal. It is possible that some methods of disposal will not be permissible because of excessive environmental effects or that the optimum disposal method may include a combination of several methods. A preliminary analysis is presented of the Hanford Nuclear Energy Center heat disposal problem to determine the best methods for disposal and any obvious limitations on the amount of heat that can be released. The analysis is based, in part, on information from an interim conceptual study, a heat sink management analysis, and a meteorological analysis

  14. Selection of heat disposal methods for a Hanford Nuclear Energy Center

    Energy Technology Data Exchange (ETDEWEB)

    Young, J.R.; Kannberg, L.D.; Ramsdell, J.V.; Rickard, W.H.; Watson, D.G.

    1976-06-01

    Selection of the best method for disposal of the waste heat from a large power generation center requires a comprehensive comparison of the costs and environmental effects. The objective is to identify the heat dissipation method with the minimum total economic and environmental cost. A 20 reactor HNEC will dissipate about 50,000 MWt of waste heat; a 40 reactor HNEC would release about 100,000 MWt. This is a much larger discharge of heat than has occurred from other concentrated industrial facilities and consequently a special analysis is required to determine the permissibility of such a large heat disposal and the best methods of disposal. It is possible that some methods of disposal will not be permissible because of excessive environmental effects or that the optimum disposal method may include a combination of several methods. A preliminary analysis is presented of the Hanford Nuclear Energy Center heat disposal problem to determine the best methods for disposal and any obvious limitations on the amount of heat that can be released. The analysis is based, in part, on information from an interim conceptual study, a heat sink management analysis, and a meteorological analysis.

  15. Momentum, heat, and mass transfer analogy for vertical hydraulic transport of inert particles

    Directory of Open Access Journals (Sweden)

    Jaćimovski Darko R.

    2014-01-01

    Full Text Available Wall-to-bed momentum, heat and mass transfer in vertical liquid-solids flow, as well as in single phase flow, were studied. The aim of this investigation was to establish the analogy among those phenomena. Also, effect of particles concentration on momentum, heat and mass transfer was studied. The experiments in hydraulic transport were performed in a 25.4 mm I.D. cooper tube equipped with a steam jacket, using spherical glass particles of 1.94 mm in diameter and water as a transport fluid. The segment of the transport tube used for mass transfer measurements was inside coated with benzoic acid. In the hydraulic transport two characteristic flow regimes were observed: turbulent and parallel particle flow regime. The transition between two characteristic regimes (γ*=0, occurs at a critical voidage ε≈0.85. The vertical two-phase flow was considered as the pseudofluid, and modified mixture-wall friction coefficient (fw and modified mixture Reynolds number (Rem were introduced for explanation of this system. Experimental data show that the wall-to-bed momentum, heat and mass transfer coefficients, in vertical flow of pseudofluid, for the turbulent regime are significantly higher than in parallel regime. Wall-to-bed, mass and heat transfer coefficients in hydraulic transport of particles were much higher then in single-phase flow for lower Reynolds numbers (Re15000, there was not significant difference. The experimental data for wall-to-bed momentum, heat and mass transfer in vertical flow of pseudofluid in parallel particle flow regime, show existing analogy among these three phenomena. [Projekat Ministarstva nauke Republike Srbije, br. 172022

  16. Heat energy from hydrogen-metal nuclear interactions

    Science.gov (United States)

    Hadjichristos, John; Gluck, Peter

    2013-11-01

    The discovery of the Fleischmann-Pons Effect in 1989, a promise of an abundant, cheap and clean energy source was premature in the sense that theoretical knowledge, relative technologies and the experimental tools necessary for understanding and for scale-up still were not available. Therefore the field, despite efforts and diversification remained quasi-stagnant, the effect (a scientific certainty) being of low intensity leading to mainstream science to reject the phenomenon and not supporting its study. Recently however, the situation has changed, a new paradigm is in statunascendi and the obstacles are systematically removed by innovative approaches. Defkalion, a Greek company (that recently moved in Canada for faster progress) has elaborated an original technology for the Ni-H system [1-3]. It is about the activation of hydrogen and creation of nuclear active nano-cavities in the metal through a multi-stage interaction, materializing some recent breakthrough announcements in nanotechnology, superconductivity, plasma physics, astrophysics and material science. A pre-industrial generator and a novel mass-spectrometry instrumentations were created. Simultaneously, a meta-theory of phenomena was sketched in collaboration with Prof. Y. Kim (Purdue U).

  17. Development of nuclear spent fuel Maritime transportation scenario

    International Nuclear Information System (INIS)

    Yoo, Min; Kang, Hyun Gook

    2014-01-01

    Spent fuel transportation of South Korea is to be conducted through near sea because it is able to ship a large amount of the spent fuel far from the public comparing to overland transportation. The maritime transportation is expected to be increased and its risk has to be assessed. For the risk assessment, this study utilizes the probabilistic safety assessment (PSA) method and the notions of the combined event. Risk assessment of maritime transportation of spent fuel is not well developed in comparison with overland transportation. For the assessment, first, the transportation scenario should be developed and categorized. Categories are assorted into the locations, release aspects and exposure aspects. This study deals with accident that happens on voyage and concentrated on ship-ship collision. The collision accident scenario is generated with event tree analysis. The scenario will be exploited for the maritime transportation risk model which includes consequence and accident probability

  18. Remodeling nuclear architecture allows efficient transport of herpesvirus capsids by diffusion.

    Science.gov (United States)

    Bosse, Jens B; Hogue, Ian B; Feric, Marina; Thiberge, Stephan Y; Sodeik, Beate; Brangwynne, Clifford P; Enquist, Lynn W

    2015-10-20

    The nuclear chromatin structure confines the movement of large macromolecular complexes to interchromatin corrals. Herpesvirus capsids of approximately 125 nm assemble in the nucleoplasm and must reach the nuclear membranes for egress. Previous studies concluded that nuclear herpesvirus capsid motility is active, directed, and based on nuclear filamentous actin, suggesting that large nuclear complexes need metabolic energy to escape nuclear entrapment. However, this hypothesis has recently been challenged. Commonly used microscopy techniques do not allow the imaging of rapid nuclear particle motility with sufficient spatiotemporal resolution. Here, we use a rotating, oblique light sheet, which we dubbed a ring-sheet, to image and track viral capsids with high temporal and spatial resolution. We do not find any evidence for directed transport. Instead, infection with different herpesviruses induced an enlargement of interchromatin domains and allowed particles to diffuse unrestricted over longer distances, thereby facilitating nuclear egress for a larger fraction of capsids.

  19. Laser-heating and Radiance Spectrometry for the Study of Nuclear Materials in Conditions Simulating a Nuclear Power Plant Accident.

    Science.gov (United States)

    Manara, Dario; Soldi, Luca; Mastromarino, Sara; Boboridis, Kostantinos; Robba, Davide; Vlahovic, Luka; Konings, Rudy

    2017-12-14

    Major and severe accidents have occurred three times in nuclear power plants (NPPs), at Three Mile Island (USA, 1979), Chernobyl (former USSR, 1986) and Fukushima (Japan, 2011). Research on the causes, dynamics, and consequences of these mishaps has been performed in a few laboratories worldwide in the last three decades. Common goals of such research activities are: the prevention of these kinds of accidents, both in existing and potential new nuclear power plants; the minimization of their eventual consequences; and ultimately, a full understanding of the real risks connected with NPPs. At the European Commission Joint Research Centre's Institute for Transuranium Elements, a laser-heating and fast radiance spectro-pyrometry facility is used for the laboratory simulation, on a small scale, of NPP core meltdown, the most common type of severe accident (SA) that can occur in a nuclear reactor as a consequence of a failure of the cooling system. This simulation tool permits fast and effective high-temperature measurements on real nuclear materials, such as plutonium and minor actinide-containing fission fuel samples. In this respect, and in its capability to produce large amount of data concerning materials under extreme conditions, the current experimental approach is certainly unique. For current and future concepts of NPP, example results are presented on the melting behavior of some different types of nuclear fuels: uranium-plutonium oxides, carbides, and nitrides. Results on the high-temperature interaction of oxide fuels with containment materials are also briefly shown.

  20. Dynamic Complexity Study of Nuclear Reactor and Process Heat Application Integration

    Energy Technology Data Exchange (ETDEWEB)

    J' Tia Patrice Taylor; David E. Shropshire

    2009-09-01

    Abstract This paper describes the key obstacles and challenges facing the integration of nuclear reactors with process heat applications as they relate to dynamic issues. The paper also presents capabilities of current modeling and analysis tools available to investigate these issues. A pragmatic approach to an analysis is developed with the ultimate objective of improving the viability of nuclear energy as a heat source for process industries. The extension of nuclear energy to process heat industries would improve energy security and aid in reduction of carbon emissions by reducing demands for foreign derived fossil fuels. The paper begins with an overview of nuclear reactors and process application for potential use in an integrated system. Reactors are evaluated against specific characteristics that determine their compatibility with process applications such as heat outlet temperature. The reactor system categories include light water, heavy water, small to medium, near term high-temperature, and far term high temperature reactors. Low temperature process systems include desalination, district heating, and tar sands and shale oil recovery. High temperature processes that support hydrogen production include steam reforming, steam cracking, hydrogen production by electrolysis, and far-term applications such as the sulfur iodine chemical process and high-temperature electrolysis. A simple static matching between complementary systems is performed; however, to gain a true appreciation for system integration complexity, time dependent dynamic analysis is required. The paper identifies critical issues arising from dynamic complexity associated with integration of systems. Operational issues include scheduling conflicts and resource allocation for heat and electricity. Additionally, economic and safety considerations that could impact the successful integration of these systems are considered. Economic issues include the cost differential arising due to an integrated

  1. Heat exchanger tubing materials for CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Taylor, G.F.

    1977-07-01

    The performance of steam generator tubing (nickel-chromium-iron alloy in NPD and nickel-copper alloy in Douglas Point and Pickering generating stations) has been outstanding and no corrosion-induced failures have occurred. The primary coolant will be allowed to boil in the 600 MW (electrical) CANDU-PHW reactors. An iron-nickel-chromium alloy has been selected for the steam generator tubing because it will result in lower radiation fields than the alloys used before. It is also more resistant than nickel-chromium-iron alloy to stress corrosion cracking in the high purity water of the primary circuit, an unlikely but conceivable hazard associated with higher operating temperatures. Austenitic alloy and ferritic-austenitic stainless steel tubing have been selected for the moderator coolers in CANDU reactors being designed and under construction. These materials will reduce the radiation fields around the moderator circuit while retaining the good resistance to corrosion in service water that has characterized the copper-nickel alloys now in use. Brass and bronze tubes in feedwater heaters and condensers have given satisfactory service but do, however, complicate corrosion control in the steam cycle and, to reduce the transport of corrosion products from the feedtrain to the steam generator, stainless steel is preferred for feedwater heaters and stainlss steel or titanium for condensers. (author)

  2. Further activities of safety culture toward nuclear transportation industry

    International Nuclear Information System (INIS)

    Machida, Y.; Shimakura, D.

    2004-01-01

    On September 30, 1999, a criticality accident occurred at the uranium processing facility of the JCO Co. Ltd. (hereinafter referred to as ''JCO'') Tokai plant, located in Tokaimura, Ibaraki Prefecture. This was an unprecedented accident in Japan's history of peaceful use of nuclear power, resulting in three workers exposed to severe radiation, two of whom died, and the evacuation and enforced indoor confinement of local residents. Nuclear power suppliers must take personal responsibility for ensuring safety. In this connection, the electric power industry, heavy electric machinery manufacturers, fuel fabricators, and nuclear power research organizations gathered together to establish the Nuclear Safety Network (NSnet) in December 1999, based on the resolve to share and improve the level of the safety culture across the entire nuclear power industry and to assure that such an accident never occurs again. NSnet serves as a link between nuclear power enterprises, research organizations, and other bodies, based on the principles of equality and reciprocity. A variety of activities are pursued, such as diffusing a safety culture, implementing mutual evaluation among members, and exchanging safety-related information. Aiming to share and improve the safety culture throughout the entire nuclear power industry, NSnet thoroughly implements the principle of safety first, while at the same time making efforts to restore trust in nuclear power

  3. In core instrumentation for online nuclear heating measurements of material testing reactor

    International Nuclear Information System (INIS)

    Reynard, C.; Andre, J.; Brun, J.; Carette, M.; Janulyte, A.; Merroun, O.; Zerega, Y.; Lyoussi, A.; Bignan, G.; Chauvin, J-P.; Fourmentel, D.; Glayse, W.; Gonnier, C.; Guimbal, P.; Iracane, D.; Villard, J.-F.

    2010-01-01

    The present work focuses on nuclear heating. This work belongs to a new advanced research program called IN-CORE which means 'Instrumentation for Nuclear radiations and Calorimetry Online in REactor' between the LCP (University of Provence-CNRS) and the CEA (French Atomic Energy Commission) - Jules Horowitz Reactor (JHR) program. This program started in September 2009 and is dedicated to the conception and the design of an innovative mobile experimental device coupling several sensors and ray detectors for on line measurements of relevant physical parameters (photonic heating, neutronic flux ...) and for an accurate parametric mapping of experimental channels in the JHR Core. The work presented below is the first step of this program and concerns a brief state of the art related to measurement methods of nuclear heating phenomena in research reactor in general and MTR in particular. A special care is given to gamma heating measurements. A first part deals with numerical codes and models. The second one presents instrumentation divided into various kinds of sensor such as calorimeter measurements and gamma ionization chamber measurements. Their basic principles, characteristics such as metrological parameters, operating mode, disadvantages/advantages, ... are discussed. (author)

  4. Comparison of the Thermal Response of Two Calorimetric Cells Dedicated to Nuclear Heating Measurements during Calibration

    International Nuclear Information System (INIS)

    Brun, J.; Reynard, C.; De-Vita, C.; Carette, M.; Muraglia, M.; Lyoussi, A.; Fourmentel, D.; Guimbal, P.; Villard, J-F.

    2013-06-01

    Nuclear heating is a key parameter which contributes to the thermal design and the quality of in-pile experiments performed in Material Testing Reactors (MTRs) for the study of nuclear materials and fuels under irradiation. Nuclear heating is typically measured in MTRs by radiometric calorimeters. However this kind of sensor has to be suited and improved in perspective of the new experimental conditions inside the channels of Jules Horowitz Reactor (JHR). In this paper, we study the responses of two non adiabatic differential calorimeter cells having the same geometric design, but different dimensions. These experimental works are carried out during a preliminary out-of-pile calibration operating procedure of these sensors which consists in simulating the sample heating by Joule effect. The influence of the imposed electrical power and of the forced cooling flow is determined on the sensor calibration curves. A more sensitive sensor leads to a quadratic calibration curve. This behavior difference of the two calorimetric configurations is explained by means of temperature and heat flux measurements performed with a new instrumented jacket. (authors)

  5. Numerical and experimental calibration of a calorimetric sample cell dedicated to nuclear heating measurements

    International Nuclear Information System (INIS)

    Brun, J.; Reynard-Carette, C.; Merroun, O.; Carette, M.; Janulyte, A.; Zerega, Y.; Andre, J.; Lyoussi, A.; Bignan, G.; Chauvin, J.P.; Fourmentel, D.; Gonnier, C.; Guimbal, P.; Malo, J.Y.; Villard, J.F.

    2012-01-01

    Online nuclear measurements inside experimental channels of material testing reactors (MTRs) are needed for experimental works (to design mock-ups) and for numerical works (input data) in order to better understanding complex phenomena occurring during the accelerated ageing of materials and the irradiation of nuclear fuels. In this paper, we focus only on one kind of measurements: nuclear heating performed by means of a radiometric calorimeter. The aims of numerical and experimental works are firstly to optimize the sensor response: in particular the sensitivity for new energy deposit ranges (new lower nuclear heating level in the reflector), and then to miniaturize and adapt this sensor for irradiation conditions in the Jules Horowitz Reactor (JHR). A calorimeter, developed previously by the CEA, is studied. It corresponds to a graphite differential calorimeter. It is used with a nonadiabatic mode called heat flow mode too. Experimental calibration of the sample cell is presented. In that case, energy deposit is simulated by Joule effect and the sample cell is inserted into a bath at a regulated temperature and controlled flow. The response of the sensor is discussed versus electrical power imposed for two flow rates. Numerical works show the influence of the gas conductivity and of specific dimensions on the cell sensitivity. (authors)

  6. Seawater desalination plant using nuclear heating reactor coupled with MED process

    International Nuclear Information System (INIS)

    Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

    2000-01-01

    A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

  7. Rapid Cooling Heat Transfer of Rod-shaped Test Specimen for Nuclear Reactor Application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Shin, Chang Hwan; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee; Koo, Yang Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Under a loss of coolant accident in pressured water nuclear reactor core, the reflood phase is associated with the emergency cooling, and in such a case, the fuel rods are quenched when water is being refilled in a reactor vessel. In the nuclear reactor core, zircaloy and SS (Stainless Steel) are widely and popularly used. Hence, the performance comparison of rapid cooling heat transfer between both materials can be meaningful. After the Fukushima accident, the hydrogen generation is considered to be one of critical issues for nuclear reactor safety. Hydrogen is generated by the corrosion reaction of zirconium alloys in nuclear reactor components. The corrosion reaction can become active with increasing the environmental temperature. Therefore, a decrease in high-temperature oxidation rate of nuclear fuel cladding should be achieved to decrease the amount of hydrogen generation under the accident condition. Recently, the researches on ATFC (Accident Tolerant Fuel Cladding) developments have been highlighted, and the chromium-coated (Cr-coated) zircaloy cladding may be considered to be one of candidates for ATFC. Thus, the investigation on the behavior of Cr-coated surface during quenching should be performed. In this paper, transient boiling heat transfer behavior of rod-shaped zircaloy and SS test specimens is studied. In addition, commercially Cr-coated SS test specimen is prepared using the plating method, and preliminarily tested for ATFC application. In this paper, transient boiling heat transfer behavior was examined using rod-shaped test specimen. For reduction in quenching time, the test specimen with small size and small heat capacity was needed. In the Cr-coated SS test specimen prepared by commercial plating method, the cracked morphology was likely to be observed on the surface after repeated quenching tests. Therefore, the optimized coating technology for ATFC application should be proposed and developed.

  8. Study on VCSEL laser heating chip in nuclear magnetic resonance gyroscope

    Science.gov (United States)

    Liang, Xiaoyang; Zhou, Binquan; Wu, Wenfeng; Jia, Yuchen; Wang, Jing

    2017-10-01

    In recent years, atomic gyroscope has become an important direction of inertial navigation. Nuclear magnetic resonance gyroscope has a stronger advantage in the miniaturization of the size. In atomic gyroscope, the lasers are indispensable devices which has an important effect on the improvement of the gyroscope performance. The frequency stability of the VCSEL lasers requires high precision control of temperature. However, the heating current of the laser will definitely bring in the magnetic field, and the sensitive device, alkali vapor cell, is very sensitive to the magnetic field, so that the metal pattern of the heating chip should be designed ingeniously to eliminate the magnetic field introduced by the heating current. In this paper, a heating chip was fabricated by MEMS process, i.e. depositing platinum on semiconductor substrates. Platinum has long been considered as a good resistance material used for measuring temperature The VCSEL laser chip is fixed in the center of the heating chip. The thermometer resistor measures the temperature of the heating chip, which can be considered as the same temperature of the VCSEL laser chip, by turning the temperature signal into voltage signal. The FPGA chip is used as a micro controller, and combined with PID control algorithm constitute a closed loop control circuit. The voltage applied to the heating resistor wire is modified to achieve the temperature control of the VCSEL laser. In this way, the laser frequency can be controlled stably and easily. Ultimately, the temperature stability can be achieved better than 100mK.

  9. Possible generation of heat from nuclear fusion in Earth's inner core.

    Science.gov (United States)

    Fukuhara, Mikio

    2016-11-23

    The cause and source of the heat released from Earth's interior have not yet been determined. Some research groups have proposed that the heat is supplied by radioactive decay or by a nuclear georeactor. Here we postulate that the generation of heat is the result of three-body nuclear fusion of deuterons confined in hexagonal FeDx core-centre crystals; the reaction rate is enhanced by the combined attraction effects of high-pressure (~364 GPa) and high-temperature (~5700 K) and by the physical catalysis of neutral pions: 2 D +  2 D +  2 D → 2 1 H +  4 He + 2  + 20.85 MeV. The possible heat generation rate can be calculated as 8.12 × 10 12  J/m 3 , based on the assumption that Earth's primitive heat supply has already been exhausted. The H and He atoms produced and the anti-neutrino are incorporated as Fe-H based alloys in the H-rich portion of inner core, are released from Earth's interior to the universe, and pass through Earth, respectively.

  10. Artificial neural network to support thermohydraulic design optimization for an advanced nuclear heat removal system

    International Nuclear Information System (INIS)

    Ridluan, Artit; Tokuhiro, Akira; Linda, Ondrej; Manic, Milos

    2009-01-01

    The U.S. Department of Energy (DOE) is leading a number of initiatives, including one known as the Next Generation Nuclear Plant (NGNP) project. One of the NGNP nuclear system concepts is the Very High Temperature (gas-cooled) Reactor (VHTR) that may be coupled to a hydrogen generating plant to support the anticipated hydrogen economy. For the NGNP, an efficient power conversion system using an Intermediate Heat Exchanger (IHX) is key to electricity and/or process heat generation (hydrogen production). Ideally, it's desirable for the IHX to be compact and thermally efficient. However, traditional heat exchanger design practices do not assure that the design parameters are optimized. As part of NGNP heat exchanger design and optimization project, this research paper thus proposes developing a recurrent-type Artificial Neural Network (ANN), the Hopfield Network (HN) model, in which the activation function is modified, as a design optimization approach to support a NGNP thermal system candidate, the Printed Circuit Heat Exchanger (PCHE). Four quadratic functions, available in literature, were used to test the presented methodology. The results computed by an artificially intelligent approach were compared to another approach, the Genetic Algorithm (GA). The results show that the HN results are close to GA in optimization of multi-variable second-order equations. (author)

  11. Transport and repair of contaminated nuclear components - liabilities and insurance

    International Nuclear Information System (INIS)

    Brunego, C.; Deprimoz, J.; Engelhard, M.

    1983-01-01

    The nuclear park has been constructed fairly recently and has not yet required large-scale maintenance efforts; however account should now be taken of the fact that periodic checks of nuclear power plants will imply systematic transfers of irradiated or contaminated materials outside the plants. In this context, the paper reviews the nuclear third party liability regime under the Paris Convention and the Euratom directives on radiation protection. It then describes the cover offered by insurance pools in several European countries. (NEA) [fr

  12. Supporting system in emergency response plan for nuclear material transport accidents

    International Nuclear Information System (INIS)

    Nakagome, Y.; Aoki, S.

    1993-01-01

    As aiming to provide the detailed information concerning nuclear material transport accidents and to supply it to the concerned organizations by an online computer, the Emergency Response Supporting System has been constructed in the Nuclear Safety Technology Center, Japan. The system consists of four subsystems and four data bases. By inputting initial information such as name of package and date of accident, one can obtain the appropriate initial response procedures and related information for the accident immediately. The system must be useful for protecting the public safety from nuclear material transport accidents. But, it is not expected that the system shall be used in future. (J.P.N.)

  13. Executive summary of safeguards systems concepts for nuclear material transportation. Final report

    International Nuclear Information System (INIS)

    Baldonado, O.C.; Kevany, M.; Rodney, D.; Pitts, D.; Mazur, M.

    1977-09-01

    The U.S. Nuclear Regulatory Commission contracted with System Development Corporation to develop integrated system concepts for the safeguard of special strategic nuclear materials (SSNM), which include plutonium, uranium 233 and uranium 235 of at least 20 percent enrichment, against malevolent action during interfacility transport. This executive summary outlines the conduct and findings of the project. The study was divided into three major subtasks: (1) The development of adversary action sequences; (2) The assessment of the vulnerability of the transport of nuclear materials to adversary action; (3) The development of conceptual safeguards system design requirements to reduce vulnerabilities

  14. Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

    International Nuclear Information System (INIS)

    Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

    1977-01-01

    The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

  15. Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

    International Nuclear Information System (INIS)

    Harvey, R.S.

    1983-01-01

    An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

  16. Review of Nuclear Heating Measurement by Calorimetry in France and USA

    Science.gov (United States)

    Reynard-Carette, C.; Kohse, G.; Brun, J.; Carette, M.; Volte, A.; Lyoussi, A.

    2018-01-01

    This paper gives a short review of sensors dedicated to measuring nuclear heating rate inside fission reactors in France and USA and especially inside Material Testing Reactors. These sensors correspond to heat flow calorimeters composed of a single calorimetric cell or of two calorimetric cells at least with a reference cell to obtain a differential calorimeter. The aim of this paper is to present the common running principle of these sensors and their own special characteristics through their design, calibration methods, and in-pile measurement techniques, and to describe multi-sensor probes including calorimeters.

  17. Power modulation for nuclear power stations supplying electric power or heat distribution networks

    International Nuclear Information System (INIS)

    Pacault, P.; Tillequin, J.

    1975-01-01

    The advantages of the thermal energy storage are presented. The heat is stored in a liquid having a high-boiling point. It is possible to use low-cost materials having an excellent stability during the temperature cycles. The application of the process to the power modulation of generating plants has been studied. Linking up heat accumulators to a nuclear power station makes it possible for the latter to supply variable amounts of electric and thermal power while keeping the reactor at constant power [fr

  18. The heat transport system and plant design for the HYLIFE-2 fusion reactor

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1990-01-01

    HYLIFE is the name given to a family of self-healing liquid-wall reactor concepts for inertial confinement fusion. This HYLIFE-II concept employs the molten salt, Flibe, for the liquid jets instead of liquid lithium used in the original HYLIFE-I study. A preliminary conceptual design study of the heat transport system and the balance of plant of the HYLIFE-II fusion power plant is described in this paper with special emphasis on a scoping study to determine the best intermediate heat exchanger geometry and flow conditions for minimum cost of electricity. 11 refs., 8 figs

  19. Finite speed heat transport in a quantum spin chain after quenched local cooling

    Science.gov (United States)

    Fries, Pascal; Hinrichsen, Haye

    2017-04-01

    We study the dynamics of an initially thermalized spin chain in the quantum XY-model, after sudden coupling to a heat bath of lower temperature at one end of the chain. In the semi-classical limit we see an exponential decay of the system-bath heatflux by exact solution of the reduced dynamics. In the full quantum description however, we numerically find the heatflux to reach intermediate plateaus where it is approximately constant—a phenomenon that we attribute to the finite speed of heat transport via spin waves.

  20. A unified framework for heat and mass transport at the atomic scale

    Science.gov (United States)

    Ponga, Mauricio; Sun, Dingyi

    2018-04-01

    We present a unified framework to simulate heat and mass transport in systems of particles. The proposed framework is based on kinematic mean field theory and uses a phenomenological master equation to compute effective transport rates between particles without the need to evaluate operators. We exploit this advantage and apply the model to simulate transport phenomena at the nanoscale. We demonstrate that, when calibrated to experimentally-measured transport coefficients, the model can accurately predict transient and steady state temperature and concentration profiles even in scenarios where the length of the device is comparable to the mean free path of the carriers. Through several example applications, we demonstrate the validity of our model for all classes of materials, including ones that, until now, would have been outside the domain of computational feasibility.